Comparison analysis of effectiveness of diagnostic methods of local coolant boiling in WWER core
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Y. Semchenkov
Abstract
The method developed by the Kurchatov Institute's WWER reactor department to detect local coolant boiling in the reactor core on the basis of the neutron noise analysis is currently being implemented at some WWER-1000 units within their regular in-core noise diagnostic systems. Given that WWER-1000 power uprating – as well as new designs being developed, such as AES-2006 and WWER-TOI characterized with higher core power and coolant temperature – would increase the probability of coolant boiling, the efficiency of the available boiling diagnostics method was to be re-evaluated. On the basis of relevant computations, it was suggested to modify the diagnostic model and introduce new parameters to improve the reliability of local boiling detection. Computations simulating in-core boiling were thus performed and showed that the improved method was quite efficient. This paper discusses the possibilities to detect – using the improved method – various in-core coolant boiling cases anticipated for WWERs.
Kurzfassung
Die im Kurchatov Institut entwickelte Methode zur Erkennung lokalen Siedens im Reaktorkern auf Basis der Analyse des Neutronenrauschens ist derzeit in einigen WWER-1000 im Rahmen der regulären Rauschmessungen im Kern implementiert. Da aktuelle Neuentwicklungen und Leistungserhöhungen bestehender Reaktoren die Wahrscheinlichkeit von lokalem Sieden erhöhen wurde die Effektivität der Methode zur Diagnose von lokalem Sieden auf ihre Anwendbarkeit für diese Bereiche erneut untersucht. Basierend auf Rechnungen wurden neue Parameter in das Diagnosemodell eingeführt, die zu einer Erhöhung der Erkennung lokalen Siedens führen sollen. Anschließende erneute Berechnungen zeigten, dass diese erweiterte Methode sehr effizient arbeitet. In diesem Beitrag werden die Möglichkeiten zur Diagnose von verschiedenen lokalen Siedefällen im Kern, wie sie innerhalb von WWER auftreten können, vorgestellt.
References
1 Semchenkov, Yu.; Milto, V.; Pinegin, A.; Shumskiy, B.: Analysis of neutron flux noise caused by coolant parameters fluctuations in WWER reactor core. Moscow, Atomnaya Energiya103 (2007) 283–286Search in Google Scholar
2 Semchenkov, Yu.; Milto, V.; Shumskiy, B.: Implementation of in-core coolant boiling diagnostic methods in WWER-1000 in-core noise diagnostic system. Moscow, Atomnaya Energiya, 105 (2008) 79–82Search in Google Scholar
3 Semchenkov, Yu.; Milto, V.; Shumskiy, B.: Method and channel to detect coolant boiling in WWER reactor core. Russian Federation Patent No 2 437 176, Inventions and Useful Models. # 35, 20.12.2011Search in Google Scholar
4 NOSTRA software package (Version 5.0). Qualification certificate # 167 of 23.12.2003. Russian Federal Nuclear and Radiation Safety Inspectorate, Moscow, 2003Search in Google Scholar
© 2014, Carl Hanser Verlag, München
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- 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
Articles in the same Issue
- 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
