Assessment of the uncertainties of COBRA sub-channel calculations by using a PWR type rod bundle and the OECD NEA UAM and the PSBT benchmarks data
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I. Panka
Abstract
The assessment of the uncertainties of COBRA-IIIC thermal-hydraulic analyses of rod bundles is performed for a 5-by-5 bundle representing a PWR fuel assembly. In the first part of the paper the modeling uncertainties are evaluated in the term of the uncertainty of the turbulent mixing factor using the OECD NEA/NRC PSBT benchmark data. After that the uncertainties of the COBRA calculations are discussed performing Monte-Carlo type statistical analyses taking into account the modeling uncertainties and other uncertainties prescribed in the OECD NEA UAM benchmark specification. Both steady-state and transient cases are investigated. The target quantities are the uncertainties of the void distribution, the moderator density, the moderator temperature and the DNBR. We will see that – beyond the uncertainties of the geometry and the boundary conditions – it is very important to take into account the modeling uncertainties in case of bundle or sub-channel thermo-hydraulic calculations.
Kurzfassung
Für ein 5 × 5 DWR-Brennelement wurden die Unsicherheiten einer thermohydraulischen Analyse mit dem Programm COBRA-IIIC bewertet. Dazu wurden zuerst basierend auf den OECD NEA/NRC PSBT Benchmarkdaten die Unsicherheiten der Modellierung durch die Unsicherheit des turbulenten Mischungsfaktors bestimmt. Darauf aufbauend wurden statistische Monte-Carlo Analysen durchgeführt, die die Modellierungsunsicherheiten und die im OECD NEA UAM Benchmark beschriebenen Unsicherheiten berechneten. Es wurden hierzu stationäre und transiente Rechnungen durchgeführt. Als Vergleichsgrößen wurden die Unsicherheiten der Dampfgehalte, der Moderatordichte und -temperatur und des DNBR herangezogen. Dabei zeigte sich, dass neben Unsicherheiten, die aus den Geometriedaten und den Randbedingungen resultieren, auch die Modellierungsunsicherheiten bei thermohydraulischen Berechnungen mit Bündel- und Unterkanalbetrachtungen berücksichtigt werden müssen.
References
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© 2014, Carl Hanser Verlag, München
Artikel in diesem Heft
- Contents/Inhalt
- Contents
- Summaries/Kurzfassungen
- Summaries
- Editorial
- Editorial
- Technical Contributions/Fachbeiträge
- Highly enriched alternatives of VVER-440 fuel assembly
- “FULL-CORE” VVER-440 calculation benchmark
- Development of approximation method to evaluate isotopic composition of burnt fuel
- Fuel assembly burnup calculations for VVER fuel assemblies with the MONTE CARLO code SERPENT
- Solution of the CB6 benchmark on VVER-440 final disposal using the Serpent reactor physics code
- Development and verification of new nodal methods in the KIKO3DMG code
- HPLWR fine mesh core analysis
- Assessment of reactor scram effectiveness based on measured worth of separate CR groups
- Engineering factors of the macrocode MOBY-DICK
- CFD investigation of flow in the MATIS-H test facility
- Investigation of the hot-channel calculation methodology in case of shroud-less assemblies
- Assessment of the uncertainties of COBRA sub-channel calculations by using a PWR type rod bundle and the OECD NEA UAM and the PSBT benchmarks data
- Comparison analysis of effectiveness of diagnostic methods of local coolant boiling in WWER core
- Sensitivity of hydrodynamic parameters' distributions in VVER-1000 reactor pressure vessel (RPV) with respect to uncertainty of the local hydraulic resistance coefficients
Artikel in diesem Heft
- Contents/Inhalt
- Contents
- Summaries/Kurzfassungen
- Summaries
- Editorial
- Editorial
- Technical Contributions/Fachbeiträge
- Highly enriched alternatives of VVER-440 fuel assembly
- “FULL-CORE” VVER-440 calculation benchmark
- Development of approximation method to evaluate isotopic composition of burnt fuel
- Fuel assembly burnup calculations for VVER fuel assemblies with the MONTE CARLO code SERPENT
- Solution of the CB6 benchmark on VVER-440 final disposal using the Serpent reactor physics code
- Development and verification of new nodal methods in the KIKO3DMG code
- HPLWR fine mesh core analysis
- Assessment of reactor scram effectiveness based on measured worth of separate CR groups
- Engineering factors of the macrocode MOBY-DICK
- CFD investigation of flow in the MATIS-H test facility
- Investigation of the hot-channel calculation methodology in case of shroud-less assemblies
- Assessment of the uncertainties of COBRA sub-channel calculations by using a PWR type rod bundle and the OECD NEA UAM and the PSBT benchmarks data
- Comparison analysis of effectiveness of diagnostic methods of local coolant boiling in WWER core
- Sensitivity of hydrodynamic parameters' distributions in VVER-1000 reactor pressure vessel (RPV) with respect to uncertainty of the local hydraulic resistance coefficients
