Fuel assembly burnup calculations for VVER fuel assemblies with the MONTE CARLO code SERPENT
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T. Lötsch
Abstract
The Monte Carlo code SERPENT (http://montecarlo.vtt.fi) is tested and validated for different fuel configurations but not for VVER fuel assemblies with Gd fuel pins and enrichment profiling. The presentation outlines the results obtained during fuel assembly burnup calculations for VVER-440 and VVER-1000 reactor types with Gd fuel pins. The calculations follow the proposal of the benchmark for VVER-440 fuel assemblies specified by Mikolas at the 14th Symposium of AER in 2004. For fuel assemblies of VVER-1000 reactor types with Gd fuel pins a benchmark by the OECD/NEA was specified in 2002. A set of data for the further verification of the few group data preparation in the framework of the VVER-1000 reactor core burnup benchmark proposed at the 19th AER Symposium in 2009 is received and analysed. The results presented show sufficient agreement with the reference values.
Kurzfassung
Das Monte Carlo Programm SERPENT (http://montecarlo.vtt.fi) ist getestet und validiert für unterschiedliche Brennstoffkonfigurationen, jedoch nicht für WWER-Brennelemente mit Gd-Brennstäben und Anreicherungsprofilierung. Die Arbeit stellt Ergebnisse dar, die bei BE-Rechnungen für WWER-440 und WWER-1000 mit Gd-Brennstäben erzielt wurden. Die Rechnungen folgen dem WWER-440-Benchmark, der von Mikolas zum 14. AER-Symposium 2004 vorgeschlagen wurde. Für WWER-1000-Brennelemente mit Gd-Brennstäben ist von der OECD/NEA 2002 ein Benchmark spezifiziert worden. Für die Verifizierung der Weniggruppendatenbereitstellung im Rahmen des auf dem 19. AER-Symposium 2009 vorgeschlagenen WWER-1000-Abbrandbenchmarks sind Daten erzeugt und analysiert worden. Die vorgestellten Ergebnisse zeigen ausreichende Übereinstimmung mit den Referenzwerten.
References
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© 2014, Carl Hanser Verlag, München
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
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
