Analyses in regulatory practice
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F. Blömeling
Abstract
The application of numerical codes is well established in the regulatory practice for nuclear power plants. Therefore, requirements for computer-aided analyses can be found in different places of the nuclear technical rules and standards. These requirements also apply for CFD analyses which reached more and more the spot light of research activities and component design, because computing capacities increased significantly in the past years. Hence, the actual status of CFD as a possible design and analysis tool in the supervising process of nuclear power plants will be discussed in this paper. The focus is set on the prerequisites which have to be fulfilled by codes and in particular by the user who wants to use CFD methods. The prerequisites will be derived directly from the existing German technical rules and standards. Finally, two examples for typical CFD applications for nuclear safety analyses are given.
Kurzfassung
Numerische Software wird in kerntechnischen Aufsichts- und Genehmigungsverfahren bei vielfältigen Fragestellungen aus den Fachgebieten Neutronenphysik, Strukturmechanik, Thermohydraulik etc. eingesetzt. Mithilfe der Software erfolgt u.a. eine Dimensionierung und Auslegung von Systemen bzw. Komponenten sowie die Simulation von Transienten und Störfällen. In der Kerntechnik werden an derartige Analysen hohe Anforderungen gestellt. Computational Fluid Dynamik (CFD) Codes wurden für die Berechnung mehrdimensionaler Strömungsvorgänge, wie sie im Reaktorkühlsystem oder im Containment auftreten, entwickelt. Für ausgewählte einphasige Strömungsphänomene liegen inzwischen umfangreiche Erfahrungen vor. Aktuell besteht ein Entwicklungs- und Validierungsbedarf bezüglich der Simulation von Mehrphasen- und Mehrkomponentenströmungen. Es wird vorgestellt, welche Möglichkeiten für CFD-Analysen heutzutage in kerntechnischen Aufsichts- und Genehmigungsverfahren bestehen. Dazu werden die grundsätzlichen Anforderungen an numerische Software diskutiert. Insbesondere wird dargestellt, welche Ansprüche das kerntechnische Regelwerk an rechnerische Analysen stellt. Abschließend werden zwei Beispiele für die Anwendung von CFD-Analysen bei kerntechnischen Fragestellungen vorgestellt.
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© 2013, Carl Hanser Verlag, München
Artikel in diesem Heft
- Contents/Inhalt
- Contents
- Summaries/Kurzfassungen
- Summaries
- Technical Contributions/Fachbeiträge
- Development of inflatable seals for the rotatable plugs of sodium cooled fast breeder reactors: a review – Part II: R&D necessities and development across the world
- Analysis of radwaste management alternatives during dismantling of Ignalina NPP systems with low level contamination
- Calculation of excitation functions for the production of Cu and Co medical isotopes
- Reliability assessment of the passive heat removal system of the VVER-1000 reactor at Kudankulam NPP
- Calculation of the neutronic behavior of minor actinides burning in a thermal research reactor using the MCNPX 2.6 code
- Steam drum level control studies of a natural circulation multi loop reactor
- Analyses in regulatory practice
- Technical Notes/Technische Mitteilungen
- Installation and measurement capacity of 3 × 592 GBq 241Am–Be neutron irradiation cell
- Calculation of the critical thickness for one-speed neutrons in a reflected slab with backward and forward scattering using modified TN method
Artikel in diesem Heft
- Contents/Inhalt
- Contents
- Summaries/Kurzfassungen
- Summaries
- Technical Contributions/Fachbeiträge
- Development of inflatable seals for the rotatable plugs of sodium cooled fast breeder reactors: a review – Part II: R&D necessities and development across the world
- Analysis of radwaste management alternatives during dismantling of Ignalina NPP systems with low level contamination
- Calculation of excitation functions for the production of Cu and Co medical isotopes
- Reliability assessment of the passive heat removal system of the VVER-1000 reactor at Kudankulam NPP
- Calculation of the neutronic behavior of minor actinides burning in a thermal research reactor using the MCNPX 2.6 code
- Steam drum level control studies of a natural circulation multi loop reactor
- Analyses in regulatory practice
- Technical Notes/Technische Mitteilungen
- Installation and measurement capacity of 3 × 592 GBq 241Am–Be neutron irradiation cell
- Calculation of the critical thickness for one-speed neutrons in a reflected slab with backward and forward scattering using modified TN method
