Modeling of the TOSQAN test facility with the lumped parameter code COCOSYS
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E. Babilas
Abstract
Hydrogen distribution in the containment during severe accidents is an important safety issue of nuclear power plants operation. The detailed knowledge of the containment thermal-hydraulics is necessary for reliable predictions of pressure, temperature and gas distribution in the containment of the NPP during postulated accidents. For this purposes, the study of the International Standard Problem (ISP 47) has been started in 1999 with the test facilities TOSQAN, MISTRA and ThAI. The paper presents the COCOSYS nodalisation of the TOSQAN test facility. The results are analysed and compared with the experimental values.
Kurzfassung
Die Verteilung des Wasserstoffs im Containment während eines schweren Unfalls ist ein wichtiges Sicherheitsproblem beim Betrieb von Kernkraftwerken. Die detaillierte Kenntnis der Thermohydraulik des Containments ist notwendig für zuverlässige Prognosen des Drucks, der Temperatur und der Gasverteilung im Containment des Kernkraftwerks wärend eines Unfalls. Deshalb wurde 1999 mit der Untersuchung des Internationalen Standardproblems (ISP 47) mit Hilfe der Testanlagen TOSQAN, MISTRA und ThAI begonnen. Die vorliegende Arbeit stellt die COCOSYS-Nodalisation der Versuchsanlage TOSQAN vor. Die Ergebnisse werden analysiert und mit experimentellen Werten verglichen.
References
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© 2006, Carl Hanser Verlag, München
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Articles in the same Issue
- Contents/Inhalt
- Contents
- Summaries/Kurzfassungen
- Summaries
- Technical Contributions/Fachbeiträge
- Minor actinide burning in a CANDU thorium reactor
- Modeling of the TOSQAN test facility with the lumped parameter code COCOSYS
- Thermal-hydraulic modeling of the onset of flow instability in MTR reactors
- 10.3139/124.100302
- 10.3139/124.100304
- Radiation transfer in inhomogeneous exponential media
- Neutron transport problems for extremely anisotropic scattering
- Milne problem for isotropic and linearly anisotropic scattering for specular and diffuse reflecting boundary conditions solved with the HN method
- ICDE-results on complete common cause failures in the light of results obtained with the POS model
- Modelling of the cumulative behaviour of Caesium and Strontium activities in nuclear fuel
- Analytical study of the closure flow inside the ETRR-2 core chimney
- Minimizing radiation exposure for the reactor staff during the dismantling of a TRIGA research reactor
