Derivation of Uncertainty Distributions for Channel Flow Rate and Fuel Critical Heat Flux Predictions for Best-Estimate Plus Uncertainty Analysis of Slow Loss-of–Reactor Power Regulation Accidents in CANDU Stations.
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| Title: | Derivation of Uncertainty Distributions for Channel Flow Rate and Fuel Critical Heat Flux Predictions for Best-Estimate Plus Uncertainty Analysis of Slow Loss-of–Reactor Power Regulation Accidents in CANDU Stations. |
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| Authors: | Parlatan, Y.1 (AUTHOR) yuksel.parlatan@alum.MIT.edu, Rogers, J.2 (AUTHOR), Koivisto, M.1 (AUTHOR) |
| Source: | Nuclear Science & Engineering. Dec2025, Vol. 199 Issue 12, p2073-2082. 10p. |
| Subjects: | CANDU reactors, Heat flux, Nuclear energy safety measures, Uncertainty (Information theory), Drag (Hydrodynamics), Prediction models |
| Abstract: | Uncertainty in the figure of merit (FOM) parameters is a central feature of the best-estimate plus uncertainty (BEPU) method, which provides insight into the analysis margins not available from other analysis methods. The FOM uncertainty distributions are formed from propagation of the variations and uncertainty distributions in the operational and modeling parameters used in simulations of a design-basis accident (DBA) scenario for a nuclear power plant. To compute an accurate FOM uncertainty distribution, it is critical to accurately quantify and account for the input parameter prediction uncertainties. The coolant flow rate through fuel channels, or more precisely, the hydraulic resistance, including the impact of two-phase flow and its distribution in the primary heat transport system and the critical heat flux (CHF) of the fuel, are two key parameters for the limiting postulated accident scenarios in a CANDU reactor for various DBAs. Prediction uncertainty distributions for these parameters can be derived by directly validating code predictions against in-reactor measurements of flow rate and experimental measurements of CHF, respectively. Such code validation circumvents the convoluted and complex approach of decomposing computer models of physical phenomena into microscopic parameters, such as interfacial mass, momentum, and heat transfer correlations, and propagation of their uncertainty distributions to obtain an overall parameter uncertainty distribution of interest. Uncertainties associated with predictions of the coolant flow rate and CHF arise due to temporal and spatial variations and uncertainties in reactor conditions, limitations of physical models and their implementation in the codes, and in the case of CHF, measurement uncertainties associated with full-scale experiments. Careful assessment of key uncertainties, specifically their magnitudes, is important for ensuring uncertainty magnitudes are not unnecessarily over- or underestimated. These uncertainties also need to be characterized properly, e.g., whether uncertainties are common to a group of reactor fuel channels or vary independently for each fuel channel. Inadequate identification or incorrect classification or characterization of uncertainties would result in an inaccurate FOM uncertainty distribution. One important focus area for this study is the distinction between apparent prediction uncertainty (the difference between code prediction and measurement) and actual prediction uncertainty (the difference between code prediction and the true value). The actual code prediction uncertainty can be calculated from the apparent code uncertainty, provided there is adequate knowledge about the measurement uncertainty. The uncertainty models developed using this approach will be used as part of the BEPU analysis for slow loss-of–reactor power regulation accidents. [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.) | |
| Database: | Engineering Source |
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| Items | – Name: Title Label: Title Group: Ti Data: Derivation of Uncertainty Distributions for Channel Flow Rate and Fuel Critical Heat Flux Predictions for Best-Estimate Plus Uncertainty Analysis of Slow Loss-of–Reactor Power Regulation Accidents in CANDU Stations. – Name: Author Label: Authors Group: Au Data: <searchLink fieldCode="AR" term="%22Parlatan%2C+Y%2E%22">Parlatan, Y.</searchLink><relatesTo>1</relatesTo> (AUTHOR)<i> yuksel.parlatan@alum.MIT.edu</i><br /><searchLink fieldCode="AR" term="%22Rogers%2C+J%2E%22">Rogers, J.</searchLink><relatesTo>2</relatesTo> (AUTHOR)<br /><searchLink fieldCode="AR" term="%22Koivisto%2C+M%2E%22">Koivisto, M.</searchLink><relatesTo>1</relatesTo> (AUTHOR) – Name: TitleSource Label: Source Group: Src Data: <searchLink fieldCode="JN" term="%22Nuclear+Science+%26+Engineering%22">Nuclear Science & Engineering</searchLink>. Dec2025, Vol. 199 Issue 12, p2073-2082. 10p. – Name: Subject Label: Subjects Group: Su Data: <searchLink fieldCode="DE" term="%22CANDU+reactors%22">CANDU reactors</searchLink><br /><searchLink fieldCode="DE" term="%22Heat+flux%22">Heat flux</searchLink><br /><searchLink fieldCode="DE" term="%22Nuclear+energy+safety+measures%22">Nuclear energy safety measures</searchLink><br /><searchLink fieldCode="DE" term="%22Uncertainty+%28Information+theory%29%22">Uncertainty (Information theory)</searchLink><br /><searchLink fieldCode="DE" term="%22Drag+%28Hydrodynamics%29%22">Drag (Hydrodynamics)</searchLink><br /><searchLink fieldCode="DE" term="%22Prediction+models%22">Prediction models</searchLink> – Name: Abstract Label: Abstract Group: Ab Data: Uncertainty in the figure of merit (FOM) parameters is a central feature of the best-estimate plus uncertainty (BEPU) method, which provides insight into the analysis margins not available from other analysis methods. The FOM uncertainty distributions are formed from propagation of the variations and uncertainty distributions in the operational and modeling parameters used in simulations of a design-basis accident (DBA) scenario for a nuclear power plant. To compute an accurate FOM uncertainty distribution, it is critical to accurately quantify and account for the input parameter prediction uncertainties. The coolant flow rate through fuel channels, or more precisely, the hydraulic resistance, including the impact of two-phase flow and its distribution in the primary heat transport system and the critical heat flux (CHF) of the fuel, are two key parameters for the limiting postulated accident scenarios in a CANDU reactor for various DBAs. Prediction uncertainty distributions for these parameters can be derived by directly validating code predictions against in-reactor measurements of flow rate and experimental measurements of CHF, respectively. Such code validation circumvents the convoluted and complex approach of decomposing computer models of physical phenomena into microscopic parameters, such as interfacial mass, momentum, and heat transfer correlations, and propagation of their uncertainty distributions to obtain an overall parameter uncertainty distribution of interest. Uncertainties associated with predictions of the coolant flow rate and CHF arise due to temporal and spatial variations and uncertainties in reactor conditions, limitations of physical models and their implementation in the codes, and in the case of CHF, measurement uncertainties associated with full-scale experiments. Careful assessment of key uncertainties, specifically their magnitudes, is important for ensuring uncertainty magnitudes are not unnecessarily over- or underestimated. These uncertainties also need to be characterized properly, e.g., whether uncertainties are common to a group of reactor fuel channels or vary independently for each fuel channel. Inadequate identification or incorrect classification or characterization of uncertainties would result in an inaccurate FOM uncertainty distribution. One important focus area for this study is the distinction between apparent prediction uncertainty (the difference between code prediction and measurement) and actual prediction uncertainty (the difference between code prediction and the true value). The actual code prediction uncertainty can be calculated from the apparent code uncertainty, provided there is adequate knowledge about the measurement uncertainty. The uncertainty models developed using this approach will be used as part of the BEPU analysis for slow loss-of–reactor power regulation accidents. [ABSTRACT FROM AUTHOR] – Name: AbstractSuppliedCopyright Label: Group: Ab 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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| RecordInfo | BibRecord: BibEntity: Identifiers: – Type: doi Value: 10.1080/00295639.2024.2411175 Languages: – Code: eng Text: English PhysicalDescription: Pagination: PageCount: 10 StartPage: 2073 Subjects: – SubjectFull: CANDU reactors Type: general – SubjectFull: Heat flux Type: general – SubjectFull: Nuclear energy safety measures Type: general – SubjectFull: Uncertainty (Information theory) Type: general – SubjectFull: Drag (Hydrodynamics) Type: general – SubjectFull: Prediction models Type: general Titles: – TitleFull: Derivation of Uncertainty Distributions for Channel Flow Rate and Fuel Critical Heat Flux Predictions for Best-Estimate Plus Uncertainty Analysis of Slow Loss-of–Reactor Power Regulation Accidents in CANDU Stations. Type: main BibRelationships: HasContributorRelationships: – PersonEntity: Name: NameFull: Parlatan, Y. – PersonEntity: Name: NameFull: Rogers, J. – PersonEntity: Name: NameFull: Koivisto, M. IsPartOfRelationships: – BibEntity: Dates: – D: 01 M: 12 Text: Dec2025 Type: published Y: 2025 Identifiers: – Type: issn-print Value: 00295639 Numbering: – Type: volume Value: 199 – Type: issue Value: 12 Titles: – TitleFull: Nuclear Science & Engineering Type: main |
| ResultId | 1 |
