The German Nuclear Society (KTG) is a non-profit scientific technical association with more than 2500 members. KTG strives for the development and dissemination of the stateof-the-art in nuclear science and technology to promote the peaceful use of nuclear energy and related disciplines. In 2009 the KTG has celebrated its 40 anniversary. The technical work of the KTG members is performed in the framework of currently 9 topical sections, which for instance jointly organize the KTG Annual Meetings, topical meetings devoted to selected subjects and several other activities. In recent years, the topical sections Reactor Safety, Thermal and Fluid Dynamics, as well as Reactor Physics and Computational Methods, have successfully established a series of meetings on “Topical Issues of Reactor Safety Research in Germany”. These meetings represent a forum for authorities, operators, technical support organizations, and vendors. A Technical Programm Committee (TPC) consisting from members of the participating topical sections, is responsible for selecting the issues to be addressed at the meeting. The acceptance and success of this series of meetings is mainly due to the current subjects dealt with. Also this forum offers the chance for a free exchange of opinions outside licensing and supervisory procedures. A balance is presented between survey papers on the background and the state of the art in the related field, specific technical presentations, and outlook papers. The participants are provided with a general overview on the area addressed and with new ideas and impetus for their future work. In order to further promote these kind of meetings the TPC has initiated co-operation with the Independent Journal for Nuclear Engineering, Energy Systems, Radiation and Radiological Protection KERNTECHNIK. For enhanced dissemination of the state-of-the-art, the key papers of the meetings are published in special issues of the journal. The 2010 meeting on “Topical Issues of Reactor Safety Research in Germany” was held at the Forschungszentrum Dresden-Rossendorf (FZD) on October 7–8 2010. It was hosted by the Institute of Safety Research (IfS) of FZD. The second day of the 2010 meeting was devoted to the ensurance of emergency core cooling (ECC) at PWRs and BWRs during loss-of-coolant accidents (LOCA) involving the release of thermal insulation material, which may lead to the clogging of the sump or the wet well strainers. The presentations were assigned to three sessions: “Overview and International Situation”, “Experiments and Model Development” and “Ensuring the Emergency Core Cooling for PWRs and BWRs”. A selection of these presentations is published in this special issue of KERNTECHNIK. In 1977, the first precursor of such a strainer clogging event occurred in BWR Gundremmingen (KRB-A), Germany. During an overfeeding transient a safety valve was damaged leading to a LOCA and almost complete release of the thermal insulation material around the break area. Some portions of the released material were collected in front of the strainers. Since KRB-A had an emergency condenser the ECC system was not started. Therefore, no coolant loaded with mineral fibers was sucked in and the fibrous material did neither clog the strainers nor enter the core. Nevertheless, the analysis of the event identified the potential of fiber entry and deposit into the core and its influence on core cooling. Because of the decommissioning of KRB-A the safety case of this event was inadequately assessed and the conclusions drawn were insufficiently distributed nationally, as well as internationally. Further events with release of insulation material and/or the clogging of strainers by mineral wool or other types of material later on occurred, for instance the Swedish NPP Bareback-2 (1992), and the US NPPs Susquehanna-2 (1988), Perry (1993), and Limerick-1 (1995). In German PWRs and BWRs several strainer modifications and backfittings were implemented over the last 10 years. These measures take into account the recommendations of the German Reactor-Safety Commission (RSK) on the robust demonstration of the effectiveness of the ECC during loss-of-coolant accidents involving the release of insulation material or other substances and on the removal of deposits from the PWR strainers by back flushing as required for PWRs in the RSK statements of July 2004 and March 2008. The RSK requirements are based on the following fundamental acceptance criteria:
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