No abstract available

[en] In the frame of a pre-study of the KNK II test program two series of experiments related to inherent safety characteristics of sodium cooled breeder reactors have been elaborated, which are one basis for the performance of experiments of the Loss Of Flow (LOF) type and the Loss Of Heat Sink (LOHS) type. Tests of this type at KNK II would -different from the earlier tests at RAPSODIE and EBR-II- provide a demonstration of the inherently safe performance in case of a significantly non-zero Doppler effect. With a suitable execution, the foreseen series of experiments allow, as explained in this report, a substantial separation of the reactivity contributions and the determination of reactivity effects, i.e. the time constants of the recouplings. The performance and evaluation of these experiments with respect to the inherent safety potential will once more underline the distinguished role of KNK II for the development of fast breeders

[en] The loss of flow accident in the fresh Mark-Ia core of the SNR-300 has been simulated with the SAS3D code. In order to cover the energetic potential, extremely pessimistic assumptions have been taken like in the covering analyses for the burnt core. The presented results show, that the possible melting energies in the fresh core reach only 50 % of those in the burnt core, since comparable conditions for fuel-coolant interactions are not given due to the high failure thresholds and the missing fission gas pressure as initiator for fuel ejection. The SAS3D analyses are also confirming the results of former studies of this accident with the CAPRI-KADIS system, which lead to almost the same results with similar boundary conditions but simpler modelling

[en] The objective of this paper is to improve the knowledge on LMFBR fuel element behaviour during protected and unprotected transients in RAPSODIE and PHENIX reactors in order to evaluate its reliability. The range of the tests performed in these reactors is sufficiently large to cover normal and also extreme off normal conditions such as fuel melting. Results of such tests allow to better establish transient design limits for reactor structural components in particular for fuel pin cladding which play a lead role in controlling the accident sequence. Three main topics are emphasized in this paper: fuel melting during slow over-power excursions; influence of the fuel element geometrical evolution on reactivity feedback effects and reactor dynamic behaviour; clad damage evaluation during a transient (essentially very severe loss of flow)

[en] The influence of fuel-coolant-interactions (FCIs) on loss of flow type core disruptive accidents has been studied for the fresh and irradiated (end of third cycle) SNR-300 Mark 1A cores. For the fresh core two accident simulations (without and with axial expansion reactivity feedback) have been considered. For these simulations it has been shown that FCIs must only be considered for the disassembly phase of the accident and these FCIs do not lead to dangerous implosive effects. The highest energy releases are obtained if FCIs are neglected. For the irradiated core the accident has been simulated without axial expansion reactivity feedback. FCIs during the disassembly phase do not lead to dangerous implosive effects. Unfortunately, for an irradiated core, pin failures and FCIs cannot be excluded during the predisassembly phase of the accident. These phenomena are discussed and several reasons are given which lead to the conclusion that these failures and FCIs during the predisassembly phase have rather limited consequences for the irradiated SNR-300 core considered

[en] The influence of other working fluids besides fuel on the conversion process of thermal into mechanical energy has been studied for primary and secondary excursions of a hypothetical loss of flow accident in the Mark-Ia core of the SNR-300. The investigations of so-called two partner systems, i.e. fuel and one additional working medium (steel, fission products or sodium) give a good understanding of the phenomena involved. Based on these studies and on some more qualitative arguments with regard to a realistic conversion process it can be concluded that the expansion work of the two-phase fuel (mesh-wise isentropic expansion to the reactor vessel cover gas volume) is a reasonable upper bound for the mechanical energy release

[en] During RAPSODIE and PHENIX reactors operation, a lot of tests have been performed in order to increase the knowledge of LMFBR fuel element behavior under normal and abnormal conditions. Three main topics are emphasized in the paper: (1) fuel melting during steady-state operation including over power conditions (two experiments with melted fuel in RAPSODIE reactor); (2) influence of fuel element geometrical evolution on reactivity feedback effects and reactor dynamic behavior (specific series of tests performed in PHENIX and RAPSODIE reactors); and (3) clad damage evaluation during abnormal transients essentially very severe loss of flow (last tests performed in RAPSODIE reactor)

[en] The SAS3D code system which has become available at Karlsruhe through the US-DOE/German BMFT information exchange agreement has been used for extensive analyses of hypothetical loss of flow accidents in the SNR-300 end of life core. This paper summarizes important assumptions and results from these analyses which have been published in full in a licensing document. Results are presented from simulations using best estimate parameters and from pessimistic bounding case studies. 11 refs

No abstract available

[en] In March and April, 1983, a series of experiments on large transients was performed on the sodium cooled fast reactor RAPSODIE at Cadarache. The main objectives of this program are an increase of existing knowledge on physical phenomena under transient conditions providing additional elements concerning verification of physical models and validation of codes relevant to analysis of incidental and accidental fast reactor behavior. The interpretation of these experiments, covering a wide range from full power and flow down to decay power and natural circulation conditions, has been applied to core dynamic behavior and transient thermohydraulics in primary and secondary circuits of the installation