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ARIANT Assessments for CANDU Analysis Using RD-14M Experiments.

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Title: ARIANT Assessments for CANDU Analysis Using RD-14M Experiments.
Authors: Cui, John1 (AUTHOR) john.cui@cnl.ca, Waddington, Geoffrey1 (AUTHOR), Wang, Shujun1 (AUTHOR), Liu, Songyu1 (AUTHOR)
Source: Nuclear Science & Engineering. 2025 Suppl 1, Vol. 199, pS898-S922. 25p.
Subjects: Thermal hydraulics, CANDU reactors, Two-phase flow, Heat transfer, Algorithms, Simulation methods & models, Nuclear reactor accidents, Nuclear reactor safety measures
Abstract: ARIANT (AlgoRIthm for Analysis of Network Thermalhydraulics) is a Canadian Nuclear Laboratories system thermal-hydraulic code for the modeling and analysis of two-phase flow and heat transfer for pressurized heavy water reactors, light water pressurized water reactors, and advanced reactor applications. This paper presents ARIANT models and simulations of RD-14M experiments, including small-break loss-of-coolant accidents, large-break loss-of-coolant accidents, loss-of-flow accidents, station blackout, and natural circulation, that are representative of accident scenarios in a CANDU reactor. ARIANT predictions of pressures, flow rates, temperatures, and void fractions are compared against the steady-state and transient data over the course of the tests. The results show that ARIANT predicted the key parameters with reasonable accuracy, as well as the overall behavior of the five transient events. These assessments support ARIANT's applicability to the corresponding CANDU design-basis accidents and demonstrate ARIANT as an alternative to existing system thermal-hydraulic codes for CANDU safety analysis. [ABSTRACT FROM AUTHOR]
Copyright of Nuclear Science & Engineering is the property of Taylor & Francis Ltd and its content may not be copied or emailed to multiple sites without the copyright holder's express written permission. Additionally, content may not be used with any artificial intelligence tools or machine learning technologies. However, users may print, download, or email articles for individual use. This abstract may be abridged. No warranty is given about the accuracy of the copy. Users should refer to the original published version of the material for the full abstract. (Copyright applies to all Abstracts.)
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  Data: ARIANT Assessments for CANDU Analysis Using RD-14M Experiments.
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  Data: <searchLink fieldCode="JN" term="%22Nuclear+Science+%26+Engineering%22">Nuclear Science & Engineering</searchLink>. 2025 Suppl 1, Vol. 199, pS898-S922. 25p.
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  Data: <searchLink fieldCode="DE" term="%22Thermal+hydraulics%22">Thermal hydraulics</searchLink><br /><searchLink fieldCode="DE" term="%22CANDU+reactors%22">CANDU reactors</searchLink><br /><searchLink fieldCode="DE" term="%22Two-phase+flow%22">Two-phase flow</searchLink><br /><searchLink fieldCode="DE" term="%22Heat+transfer%22">Heat transfer</searchLink><br /><searchLink fieldCode="DE" term="%22Algorithms%22">Algorithms</searchLink><br /><searchLink fieldCode="DE" term="%22Simulation+methods+%26+models%22">Simulation methods & models</searchLink><br /><searchLink fieldCode="DE" term="%22Nuclear+reactor+accidents%22">Nuclear reactor accidents</searchLink><br /><searchLink fieldCode="DE" term="%22Nuclear+reactor+safety+measures%22">Nuclear reactor safety measures</searchLink>
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  Data: ARIANT (AlgoRIthm for Analysis of Network Thermalhydraulics) is a Canadian Nuclear Laboratories system thermal-hydraulic code for the modeling and analysis of two-phase flow and heat transfer for pressurized heavy water reactors, light water pressurized water reactors, and advanced reactor applications. This paper presents ARIANT models and simulations of RD-14M experiments, including small-break loss-of-coolant accidents, large-break loss-of-coolant accidents, loss-of-flow accidents, station blackout, and natural circulation, that are representative of accident scenarios in a CANDU reactor. ARIANT predictions of pressures, flow rates, temperatures, and void fractions are compared against the steady-state and transient data over the course of the tests. The results show that ARIANT predicted the key parameters with reasonable accuracy, as well as the overall behavior of the five transient events. These assessments support ARIANT's applicability to the corresponding CANDU design-basis accidents and demonstrate ARIANT as an alternative to existing system thermal-hydraulic codes for CANDU safety analysis. [ABSTRACT FROM AUTHOR]
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  Data: <i>Copyright of Nuclear Science & Engineering is the property of Taylor & Francis Ltd and its content may not be copied or emailed to multiple sites without the copyright holder's express written permission. Additionally, content may not be used with any artificial intelligence tools or machine learning technologies. However, users may print, download, or email articles for individual use. This abstract may be abridged. No warranty is given about the accuracy of the copy. Users should refer to the original published version of the material for the full abstract.</i> (Copyright applies to all Abstracts.)
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        Value: 10.1080/00295639.2024.2380628
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      – Code: eng
        Text: English
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        StartPage: S898
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        Type: general
      – SubjectFull: CANDU reactors
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      – SubjectFull: Two-phase flow
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      – SubjectFull: Heat transfer
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      – SubjectFull: Algorithms
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      – SubjectFull: Simulation methods & models
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      – SubjectFull: Nuclear reactor accidents
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      – SubjectFull: Nuclear reactor safety measures
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      – TitleFull: ARIANT Assessments for CANDU Analysis Using RD-14M Experiments.
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              Text: 2025 Suppl 1
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