[en] An experiment and analysis performed on an inverted U-tube steam condenser (similar to a steam generator) to determine the modes of flow that can exist as the rate of steam flow into the condenser is reduced. The condenser consisted of four largely glass tubes approximately 1.2 m in height connected to a common inlet plenum and a common exit plenum. Heat transfer and flow rate measurements, as well as visual observations were made. 8 refs

[en] Small scale experiments were performed to investigate the condensation process and hydrodynamic pressure oscillations when steam was discharged into a subcooled water pool. The dynamic behavior of subsonic jets, differed from that of sonic jets. The interfacial motion of a subsonic jet was periodic, composed by three intervals: bubble growth, bubble translation, and bubble separation (necking). 15 refs

[en] Testing has been performed using two 1/6 scale models of the Clinch River Breeder Reactor Plant Steam Generators to simulate fluid temperature fluctuations in the sodium. Water was used as the working fluid and the models were fabricated from Plexiglas to facilitate flow visualization. Hydraulic scaling was achieved through Richardson Number and Euler Number similarity. 4 refs

[en] KODEX is a modular computer code to analyze the consequences of an anticipated steam explosion during a core melt accident. KODEX consists of modules which determine the core meltdown, the fragmentation, the energy flow from the fragmented melt to the water around it, the pressure-time history within the RPV, and the RPV-wall response to that impulse. The highest mechanical loads on the RPV occur at the lower and the upper plenum. A loss of the integrity in the cylindrical part of the RPV can be excluded. Steam explosions with high amounts of finely fragmented CORIUM-melts are expected to result in dynamic loads below the failure criterion of the RPV. For final qualification of the maximum CORIUM mass participating in a steam explosion, the RPV is able to withstand, a more detailed knowledge of the global meltdown and slumping procedure until failure of the core support structure, of the fragmentation phenomena and a consideration of the RPV-internals is needed. To give results about the risk potential of steam explosions the sensitive parameters described have to be varied. 4 refs

[en] Emergency core cooling systems are designed to maintain fuel cladding temperature below prescribed limits even under a wide range of hypothetical loss-of-coolant accidents. The effects of the ECC during a design basis accident on the system response were investigated in the BWR system simulator under a cooperative experimental research program sponsored by the Nuclear Regulatory Commission, the Electric Power Research Institute, and the General Electric Company. 14 refs

[en] Striping phenomena at the interface of thermal stratification is applied to integrity of components and piping of LWR and FBR. An experimental study was conducted to investigate the thermal behavior of density interface. This test result suggests that it is too conservative to use the temperature fluctuation at the stratified interface along the centerline of the pipe for the realistic structural evaluation. 7 refs

[en] The results of CORECT II Experiment, which is a thermal interaction experiment of the shock tube type between uranium dioxide and sodium, are presented. The influence of the interaction zone constraint is emphasized and parametric calculations are compared with the experimental results. 4 refs

[en] The characteristic feature of these experiments is the controlled ''fine fragmentation'' and mixing without preceding coarse pre-mixing, quasi one-dimensional constraint which allows the measurement of the vapor explosion work as a function of time and the determination of variables affecting the vapor explosion. For a real situation these studies represent the phase when coarse pre-mixing is established and the vapor explosion is triggered by spontaneously appearing pressure shocks or externally introduced ones. For the present experiments one can interpret the kinetic impact as trigger causing fine fragmentation and mixing. 12 refs

[en] A steady state thermal detonation model including rapid heat transfer from the melt to the coolant induced by hydrodynamic fragmentation is used to interpret tin-water and aluminum-water experiments. Taking best estimate model input data for the initial vapor fraction, the drag coefficient and the dimensionless breakup time, a maximum pressure of 1.47 MPa in the wave is calculated which is in good agreement with the experimental results. The aluminum-water reactions can only be explained if adapted model input data are used. Because of this discrepancy the influence of important initial parameters is investigated. The model is applied to reactor materials in order to get an idea on the pressure behavior and the rapid release of thermal energy caused by large scale vapor explosions within a hypothetical core melt down accident. 16 refs

[en] Evaluations of the conditions associated with the liquid-liquid film boiling mixing of molten core material and water demonstrate that 1) mixing on a time scale comparable to the explosion requires more energy than the explosion itself, and 2) a mixing condition which evolves over a second or longer generates sufficient steam in a film boiling condition to disperse any overlying liquid and also separate the constituents. Fragmentation in a film boiling state can drive away the water and destroy a potentially exposive configuration. A study of the triggering mechanism from experimental evidence suggests that LWR configurations would be triggered by contact of the molten debris with a solid, wetted surface. Evaluations lead to the conclusion that only a limited amount of core material could be involved in an explosive interaction and such an event would not threaten the reactor pressure vessel. 23 refs