Enhancement of Corium Coolability with CRGTs in the Lower Head of a BWR
2004 (English)In: Proceedings of the 2004 International Congress on Advances in Nuclear Power Plants, ICAPP'04, American Nuclear Society, 2004, 950-962 p.Conference paper (Refereed)
A series of experiments were preformed in the POMECO (Porous Media Coolability) and the COMECO (Corium Melt Coolability) test facilities at the Nuclear Power Safety Division of the Royal Institute of Technology, Stockholm. During the experiments particulate debris beds and molten pools were cooled by establishing a water layer above them. The main aim of the experiments was to investigate the additional coolability capacity offered by the Control Rod Guide Tubes (CRGTs) in the lower head of the Reactor Pressure Vessel (RPV) of a Boiling Water Reactor (BWR). Each CRGT has a substantial heat transfer area and there is large number of these tubes in the BWR lower head. The coolant is supplied to the RPV via the CRGTs during the normal reactor operation and this coolant flow could be maintained during a severe accident. The primary objective of the experimental program was to obtain data on the heat removal capacity, offered by a CRGT. This paper presents results of the experiments.
Place, publisher, year, edition, pages
American Nuclear Society, 2004. 950-962 p.
IdentifiersURN: urn:nbn:se:kth:diva-53590ScopusID: 2-s2.0-14844328165ISBN: 0-89448-680-2OAI: oai:DiVA.org:kth-53590DiVA: diva2:470388
the 2004 International Congress on Advances in Nuclear Power Plants, ICAPP'04. Pittsburgh, PA. 13 June 2004 - 17 June 2004
QC 201201032011-12-282011-12-282012-01-03Bibliographically approved