No abstract available

No abstract available

No abstract available

[en] The reactor study is well underway seen from a broad spectrum of disciplines and technology. The objective power output with a high efficiency direct cycle power conversion unit remains promising after compiling the first critical analysis of the core and the power conversion unit. The stability and controllability of the system are demonstrated by the engineering simulator. The main system and components are basically specified for costing purposes. A first plant layout has been completed demonstrating the positions of main components, personnel movement, installation methods for large components, etc. A cryptic report style presentation includes study objectives, indicating guiding documents, giving an overview of design and analyses work done as well as a few sketches and diagram are included in this paper. Most of these sketches and diagrams are small replicas of large drawings and are therefore not readable but can be used as references. (author)

[en] The system parameters chosen to optimise the thermal efficiency of the Eskom PBMR whilst maintaining component simplicity is discussed. Power Conversion Unit components, which are now at a preliminary design stage comprise a precooler, two turbo units consisting of a turbine driven compressor, recuperator and a power turbine, driving an alternator. Design aspects of every component is mentioned and the inventory method of poorer control is explained with reference to start-up and and shut-down events, the system an effective load following device, down to 4% of full power. Application of the same design principles for HTGRs smaller than 25 MWe is discussed. (author)

[en] Cooling of the Pebble Bed Nuclear Reactor under evaluation in South Africa is primarily effected by the flow of helium through the cavity which contains the nuclear fuel. However, apart from this, a certain amount of heat flows from the reactor cavity, through the graphite barrel and reactor vessel to the containment building and ultimately to the environment During normal operation this passive heat loss represents approximately 1MW for a 100MW reactor and constitutes an undesirable loss of power. In the event of a shutdown or loss of main coolant, however, this passive heat removal is relied upon to remove the decay heat from the core. A study was initiated to simulate the process of this heat removal to provide an indication of the maximum vessel temperature and power transfers after shutdown. However, there is a lack of precise data indicating values for thermal conductivity, heat transfer coefficients, heat capacities or even densities. This paper describes the assumptions made and the manner in which these data were estimated so as to provide what is hoped to be a reasonably accurate estimate of the behaviour of the passive heat removal process. (author)

[en] The paper describes the approach taken to establish the maintenance policy for safety-related systems to comply with pre-determined reliability criteria. The design safety guidelines for the Advanced Gas Cooled Reactors require the system unreliability not to be higher than 10E-3 for credible initiating events. For the maintenance tasks, the procedures are developed within a ''Maintenance Inspection and Test Schedule'' document. (author). 4 refs

[en] This patent describes a gas cooled nuclear fuel element. It comprises a cylindrical tube having one open end; nonporous nuclear fuel disks stacked inside the cylindrical tube so as to define an annular space between the disks and the tube; the nuclear fuel disks each having an axial channel there-through and grooves to provide a path for coolant flow on a surface of each disk that extend radially between the axial channel and the outer edge of the disks; and an upper and a lower end fitting in the cylindrical tube that support the nuclear fuel disks in the tube, the lower end fitting having a central bore in coaxial alignment with the axial channel in the fuel disks

[en] Based on the analysis of results of calculations conducted according to the GAVROSh program, the comparison of neutron-physical and thermal-hydraulic characteristics of the VGM reactor, when operating according to the principles of once or multiple spherical fuel element passage through the core, is considered. It is shown that under an increase of fuel element circulation ratio the neutron-physical characteristics of the reactor are improved. An increase of the ratio of spherical fuel element circulation through the core leads to reduction of accidental reactivity effect conditioned by water ingress to the primary circuit

[en] Results of analysis of fuel power and temperature change in the VG-400 and VGM reactors in the case of design basis accidents accompanied by possitive reactivity insertion, namely, under moiture ingress to the core as a result of steam generator steam collector rupture and under the regulating rod ejection, are presented. The analysis is conducted under design basis accident occurrence, i.e. under emergency protection actuation and in a hypothetic situation when a complete failure of emergency protection takes place. It is shown that accidents with moisture spreading into the core under reactor operation at a decreased flow rate level, appear to be the most severe ones from tha viewpoint of reactivity and its growth rate. A conclusion is made that even in the case of a hypothetical situation progress under the failure of emergency protection the maximum fuel temperature does not exceed the limiting admissible one (1600 deg C)