[en] The computer programme BLOW-3A describes sodium boiling phenomena in subassemblies of fast breeder reactors as well as in in-pile or out-of-pile experiments simulating different failure conditions. This report presents a complete documentation of the code from three main viewpoints: the theoretical foundations of the programme are first described with particular reference to the most recent developments; the structure of the programme is then explained in all details necessary for the user to get a rapid acquaintance with it; eventually several examples of the programme validation are discussed thus enabling the reader to acquire a full picture of the possible applications of the code and at the same time to know its validity range. (orig.)

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[en] In the development of the accelerator driven systems XT-ADS and EFIT lead-bismuth eutectic alloy (LBE) and lead are foreseen as coolant respectively. For calculation of the heat transfer from the fuel rod surfaces to the coolant correlations are necessary. Existing correlations for triangular and square fuel pin arrangements are reviewed with respect to their experimental qualification. Effects on the heat transfer in reactor application are addressed, which could alter the correlations due to deviating boundary conditions. These are the influence of spacers and consequences due to the axial power profile. (orig.)

[en] The current ADS (Accelerator Driven System) design is based on a fast core. Therefore it is quite natural to adapt the SAS4A code to an ADS simulation for transient analyses. The current study shows that the point kinetic model in SAS4A code enables the activation of an external source with relative few modifications. In addition, localized reactivity feedback coefficients and the power distribution in an ADS must be known for a SAS4A transient core simulation. The use of perturbation theory for ADS, successfully used in homogeneous problems, is still not resolved conclusively since some parameters are undefined, in particular the adjoint weighting function and the adjoint definition of the external source. The use of perturbation theory for the calculation of localized reactivity coefficients for ADS seems therefore not applicable. These reactivity coefficients can also be determined by means of successive criticality calculations. This can be done by determining differences of multiplicity factor depending on changes in local core materials properties against the original state. The enhanced computational time requirements are acceptable. The codes package KAPROS was initially used in the current study to investigate the applicability of perturbation theory for ADS, and to demonstrate the differences between source free systems (using D3E/D3D codes) and ADS. In particular the R-Z option of the DIXY code which allows for correct multiplicity factor calculation, was used for ADS simulation. Subsequently, the CITATION code was used to calculate the reactivity state for various core conditions. This code allows for the three dimensions hexagonal representation of any ADS configuration. The results of these calculations are then used to calculate all the relevant local reactivity perturbations. The collapsed multi-group cross-section sets, which serve as input to the CITATION code, were determined with the KAPROS code system. The three sources ADS configuration, which was selected as reference case in this study, can be modified to any desired configuration dependent on the particular requirements, such as transmutations optimization or some other relevant criteria. (orig.)

[en] The fuel relocations from the CABRI-1 experiments with irradiated fuel that had been evaluated from the hodoscope measurements were used together with fuel reactivity worth distributions from the SNR-300 to estimate the reactivity effect which these motions would have if they occurred in SNR-300 at the same relative distance to the peak power as in CABRI. The procedure for the reactivity evaluation is outlined including the assumptions made for fuel mass conservation. The results show that the initial fuel motion yields always negative reactivities. They also document the mechanism for a temporary reactivity increase by in-pin fuel flow in some transient overpower tests. This mechanism, however, never dominates, because material accumulates always sufficiently above the peak power point. Thus, the late autocatalytic amplifications of voiding induced power excursions by compactive in-pin fuel flow, that had been simulated in bounding loss of flow analyses for SNR-300, have no basis at all when considering the results from the CABRI-1 experiments

[en] Experimental and theoretical analyses of sodium boiling behavior performed at the Nuclear Research Center Karlsruhe have led to the development of the computer program BLOW-3A which describes sodium boiling phenomena under accident conditions in one-dimensional geometry by means of a heterogeneous multi-bubble slug ejection model. The main characteristics of the code are explained with particular emphasis on the analysis of the transient fuel behavior and on the methods used for the numerical solution of the set of equations describing the separated phases in the two-phase flow model. Some representative results from the theoretical interpretation of single-pin and 7-pin in-pile and out-of-pile experiments are shown to provide a validation of the theoretical basis of the computer program. Future developments of the model to more-dimensional description of sodium boiling in bundle geometry are presented

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[en] The analyses of in-vessel accident sequences of LWRs aim at answering the following questions: - What is the mass of hydrogen released during a core degradation process with and without emergency cooling, respectively; - How is the material redistribution in the core up to the eventual formation of a melt pool; which failure modes of the crust of solidified core material are possible leading to the release of melt from the pool into the lower plenum; - What is the temperature history of the RPV wall during interaction with the core melt in the lower plenum. The SCDAP/RELAP5 computer code is used on a IBM RISC 6000. In a first step towards reactor case studies, out-of-pile experiments CORA-7 and CORA-13 have been calculated with SCDAP/RELAP5/ver7AG. The comparison with experimental results shows that transient fuel rod temperatures in the upper part of the CORA bundles generally agree rather well, whereas significant differences in the axial temperature profiles are found. Moreover, the calculation of CORA-13 indicated that the complex physical process of reflood can be calculated by SCDAP/RELAP5 without numerical difficulties. In parallel, preparations for the calculation of the core behavior of the EPR (European Pressurized Water Reactor) have been started in cooperation with Siemens/KWU). (orig.)

[en] In the context of consequence evaluation of LOF-accidents for SNR-300 the potential for recriticality events during core meltdown and the molten core material redistribution in the reactor vessel have been analyzed. Parametric investigations of both items are presented. 13 refs