Investigation of the hot-channel calculation methodology in case of shroud-less assemblies
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Á. Tóta
Abstract
The fulfillment of the safety analysis acceptance criteria and the normal operation safety related limitations are usually evaluated by using separate hot-channel or/and hot-assembly thermal hydraulic calculations, especially for the closed VVER-440 assemblies. However, the shroud-less assemblies of the expected new NPP units in Hungary or even those of some recently foreseen VVER-440 are raising several questions; like what is the role of the coolant mixing between the neighboring assemblies and can the closed assembly approximation be regarded as conservative for DBA analyses. Focusing an anticipated transient without scram (ATWS) event, the paper evaluates the thermal hydraulic calculation methodology of the shroud-less assemblies by supplementing the computational domain step by step with the closest pins and subchannels of the neighboring assemblies. We concluded that assembly-wise coolant cross-flows and significant changes in the hot-channel DNBR can appear in case of shroud-less assemblies.
Kurzfassung
Zum Nachweis der Erfüllung der Kriterien bei Sicherheitsanalysen und der Einhaltung der Grenzwerte beim Normalbetrieb werden thermohydraulische Berechnungen mit Heißkanal- bzw. sog. heißen Brennelementen durchgeführt, insbesondere bei geschlossenen Brennelementkästen. Offene Brennelemente, wie sie in den voraussichtlichen neuen KKW in Ungarn oder ähnlichen aktuellen WWER-440 vorgesehen sind, werfen einige Fragen auf: Welchen Einfluss hat die Kühlmittelvermischung zwischen benachbarten Brennelementen und kann bei Auslegungsstörfällen die Annahme geschlossener Brennelemente als konservativ angesehen werden? In diesem Beitrag wird am Beispiel eines ATWS Ereignisses die Heißkanalmethode für offene Brennelemente angewendet unter Berücksichtigung der benachbarten Brennstäbe und Unterkanäle der Nachbarbrennelemente. Dabei zeigt sich, dass bei offenen Brennelementen Querströmungen zwischen den Brennelementen und Änderungen des Verhaltens der Heißkanäle in den Rechnungen 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
