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[en] Safety of the various types of nuclear reactor (light water, heavy water, gas-cooled) is compared. Failures of various systems and components (e.g. cooling-medium failure) are analysed. (H.E.G.)

No abstract available

[en] The invention relates to nuclear reactor fuel element assemblies in which the coolant employed is light or heavy water. The assembly is one in which nuclear fuel rods are supported in a cluster and spaced apart by transverse grids so that the spaces between the rods form sub-channels through which main reactor coolant is passed. In order to provide auxilliary cooling it is proposed that auxilliary cooling conduits be located amongst the cluster of fuel rods. These conduits have coolant ejection holes in the wall arranged circumferentially in a predetermined pattern which is determined by the position of the conduit in the cluster and by the axial distance of the ejection hole along the conduit. (J.R.)

[en] ASTEC, jointly developed by IRSN and GRS, is considered as the European reference code since it capitalizes knowledge from the European research on the domain. The CESAM project aims at its enhancement and extension for use in severe accident management (SAM) analysis of the nuclear power plants (NPP) of Generation II-III presently under operation or foreseen in near future in Europe, spent fuel pools included. Within the CESAM project 3 main types of research activities are performed: -) further validation of ASTEC models important for SAM, in particular for the phenomena being of importance in the Fukushima-Daichi accidents, such as reflooding of degraded cores, pool scrubbing, hydrogen combustion, or spent fuel pools behaviour; -) modelling improvements, especially for BWR or based on the feedback of validation tasks; and -) ASTEC applications to severe accident scenarios in European NPPs in order to assess prevention and mitigation measures. An important step will be reached with the next major ASTEC V2.1 version planned to be delivered in the first part of 2015. Its main improvements will concern the possibility to simulate in details the core degradation of BWR and PHWR and a model of reflooding of severely degraded cores. A new user-friendly Graphic User Interface will be available for plant analyses

[en] Post-dryout heat transfer from fuel elements is important in the safety analysis of water-cooled nuclear reactors. Limited availability of relevant data has resulted in much uncertainty in the prediction of post-dryout fuel surface temperatures. Recent studies have permitted the development of a better prediction technique based on non-equilibrium effects. (author)

[en] A nuclear fuel element is described, comprising a cluster of nuclear fuel rods seperated by transverse grids so as to define interspaces for the axial flow of reactor coolant. At least one of the spaces is occupied by an axially extending auxilliary coolant, conduit with lateral holes through which an auxilliary coolant is sprayed into the cluster. The invention in particular describes a deflector means extending from the transverse grid into a position in front of the holes for deflecting the auxilliary coolant spray on to parts of the fuel rods inaccessible to auxilliary coolant. (J.R.)

[en] The description is given of a fuel assembly for a nuclear reactor moderated and cooled by pressurized water, comprising a nuclear fuel, multiple closely packed fuel pins, their longitudinal axes being parallel, each one having a generally tubular cladding carrying the fuel, characterized by the fact that the surface of the pin cladding has at least one longitudinal protrusion, rib, rim or fin and that means are provided for assembling the fins to juxtaposed pins and forming an integral assembly