“FULL-CORE” VVER-440 calculation benchmark
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V. Krýsl
Abstract
Because of the difficulties with experimental validation of power distribution predicted by macro-code on the pin by pin level we decided to prepare a calculation benchmark named “FULL-CORE” VVER-440. This benchmark is a two-dimensional (2D) calculation benchmark based on the VVER-440 reactor core cold state geometry with taking into account the geometry of explicit radial reflector. The main task of this benchmark is to test the pin by pin power distribution in fuel assemblies predicted by macro-codes that are used for neutron-physics calculations especially for VVER-440 reactors. The proposal of this benchmark was presented at the 21st Symposium of AER in 2011. The reference solution has been calculated by MCNP code using Monte Carlo method and the results have been published in the AER community. The results of reference calculation were presented at the 22nd Symposium of AER in 2012. In this paper we will compare the available macro-codes results of this calculation benchmark.
Kurzfassung
Da eine Leistungsverteilung, die mit einem Macro-Code Stab mit einer stabweisen Auflösung berechnet wird, nur sehr schwer experimentell validiert warden kann, wurde der WWER-440 Berechnungsbenchmark “Full Core” definiert. Dies ist ein 2D Benchmark basierend auf dem kalten Zustand eines WWER-440 Reaktorkerns unter Berücksichtigung der Geometrie des radialen Reflektors. Ziel des Benchmarks war der Vergleich der mit Macro-Codes berechneten stabweisen Leistungsverteilung. Dieser Benchmark wurde auf dem 21. AER-Symposium in 2011 vorgestellt. Als Referenzlösung wurde eine Rechnung mit dem Programm MCNP basierend auf der Monte Carlo Methode auf dem 22. AER-Symposium in 2012 veröffentlicht. In diesem Beitrag wird ein Vergleich der verschiedenen Benchmarkrechnungen vorgestellt.
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
