[en] The initial events considered were a long-term outage of the main coolant system, power failure, primary coolant circuit leakages, and steam generator heating tube, leakage. The relevant consequences are practically nothing but the remobilisation of fission products which have deposited at the primary coolant circuit surfaces during normal operation. The most frequent case was response of the primary coolant circuit pressure relief valve. Pressure drops in the primary coolant circuit may also be a consequence of steam generator heating tube failures causing an incident. (DG)

[en] The accidental risk emanating from the HTR-Modul is decisively determined by the fission products accumulating outside the core in the primary loop during normal operation; release of these proceeds only under conditions created by the incidents of water ingress, or depressurization. The trend analysis therefore investigates the fission products depositing within the primary loop and their release and conditions of transport for the case of steam generator leakage, or leak in the primary loop pipework. The investigation is made for the radiologically important fission products cesium, strontium, and iodine. (orig./DG)

[en] The volume of steam ingress depends inter alia from the size of the leak. The design-basis accident is defined to be severence type break of a steam generator pipe with a leak size of double pipe section. This leak size and a smaller one is taken as the basis for assessment of pressure buildup in the primary circuit. In addition, pressure buildup through ingressing steam is calculated for the case of complete steam generator piping severence at the superheater side. This is equivalent to the case of the tubeplate connection cylinder being torn off, which has been chosen as a design-basis accident phenomenon in earlier HTR design concepts. (orig./DG)

[en] Due to the fact that the fuel element temperature is low in the HTR-Modul, even maximum values in case of an incident, fission product release from the pebble-bed core is expected to be minimal also in case of failure of the forced circulation cooling system. Calculations are made using the FRESCO code that allows calculation of fission product release from the fuel sphere under limiting conditions during normal irradiation, and for incident conditions. The data describing the contamination of the primary circuit including the gas purification system are rough assessments. (DG)

[en] The study of the primary coolant loop behaviour concentrates on the case of failure of circulator cut-off equipment following failure of the steam generator feeding line, as well as on the natural convection in the primary circuit under pressurized conditions, with the isolating valve of the circulator in 'open' position. Calculated or assessed data are presented relating to the temperature buildup in the support structures under incident conditions, to the strains affecting the area of close contact of the claws and the core cylinder, and to circulator system behaviour in case of failure of the steam generator feeding line and of the circulator cut-off system. (DG)

[en] The study investigates the accident of rupture of the sphere discharge pipe and determines the reactivity data for two different configurations of equal volume, comparing the surface area. The data are obtained by two-dimensional diffusion calculations using the ASTERIX-II code. The results show a reactivity decrease of 15.3% and 42.4%, respectively. (DG)

[en] The accident consequences and risk assessments presented here take as a basis the results of previous studies made on the HTR-modul accidental behaviour. The event scenarios are used to assess the resulting release of fission products (type, amounts, frequency), which is classified into five release categories. Due to the design characteristics of the reactor, the core heat-up accident is not the dominating case to be studied. The investigations concentrate on events entailing release of certain amounts of fission products previously deposited on metallic surfaces of the system. The calculations made with the UFOMOD/B3 code show that for none of the release categories defined occurrence of immediate radiation effects is to be expected, even if no evacuation measures are taken. (orig./DG)

[en] The paper includes the reliability assessment for the surface cooling system, the gas purification system, emergency power facility, off-gas filter system and sub-atmospheric pressure holding system, the steam generator relief and steam generator shutdown system, as well as pressure suppression within the primary coolant circuit and the confinement. Among the issues to be evaluated there are also: Non-availability of the primary coolant circuit upon startup for after-heat removal. Event frequency and non-availability data are stated in tables. (DG)

[en] Completeness of risk assessment models is a control problem that becomes particularly clear through the safety analysis of the HTR-modul as with this reactor, the serious core heat-up accidents with the relevant scenarios that encompass all other type of accidental release are left out of consideration. In addition, a trend analysis that has to present data within a restricted period of time cannot be complete in this sense. The results obtained to far indicate: Only extreme external effects, or purposefully planned action of a sabotage nature are conceivalble to create accidents that would radically falsify the accident scenario taken as a basis in this study. Such effects and action are of the type that hitherto has been dealt with in all risk analyses (still) quite unsatisfactorily from the methodological point of view. Considering this fact, the trend analysis presented can be called complete. (orig./DG)

[en] 9 individual papers deal with the following subjects: Plant description, analyses of systems and courses of events, system reliability, primary coolant circuit thermodynamics and component load, processes in the event of a steam generator leakage, fission product release from the HTR-modul reactor core, fission product release from the plant due to an incident, reactivity behaviour of fuel spheres in the discharge room of the HTR-modul, consequences of an accident and related risks; description of a complete risk scenario. (DG)