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The course of severe FBR accidents, such as a single subassembly meltdown or disruptive whole-core accidents, is predominantly determined by the material motions and redistributions. In experiments using thermite to simulate the fuel and the chemical energy to simulate nuclear heat production the material redistribution was investigated in bundles undergoing meltdown. The use of X-ray cinematography allowed the observation of transient material movements. The fine dispersion of the material in the initiating phase, when driven by a steep pressure gradient, led to downstream material expulsion, which has been roughly quantified in the tests. Subsequent material flow resulted in blockage build-up in the simulated breeding zones. Three modes of hexcan failure were identified. The results of the SIMBATH test series (SIMBATH, simulation experiments in fuel element mock-ups with thermite) will be compared with those of in-pile experiments.

Under the SIMBATH programme the physical phenomena of transient material movement and relocation during severe LMFBR accidents are investigated out-of-pile. In most of the SIMBATH bundle experiments a failure of the wrapper was observed. From the safety point of view this has implications on the issue of propagation. By openings into the inter-subassembly gaps pressure relief and material release are possible. From the development of failure, based on measurements made during the simulation tests, and from post-experiment investigations three types of failure mode have been identified: • - Melt-through of the wrapper wall by a jet of hot material from a failing pin. This happened very early during the test. Sodium boiling in the annular bypass prior to failure has not been detected. • - Melt-through in the simulated fuel region by severe ablation due to local crust instability combined with intense heat input from the flowing melt. • - Melt-through in the simulated breeding regions close to blockages. This failure mode was always observed together with sodium gross boiling in the annular channel, i.e. reduced cooling of the wrapper wall. No mechanical failure was detected as a result of the stress concentration in the corners of the hexcan walls. The influence of the internal overpressure is restricted mainly to final break-through after severe ablation and drives the material motions after wrapper failure; it does not control wrapper wall failure in these experiments.

Much attention has been given in LMFBR safety analysis to cooling disturbances caused by local blockages within a fuel subassembly. Such blockages are generally considered to be more probable in gridded fuel pin clusters which present the possibility for solid particles in the coolant to be trapped at grids to form a radially extending flow obstruction. The temperature distribution produced in the region of impaired cooling has been studied in water and sodium experiments in pin bundles of various sizes. The experimental work at KfK on local cooling disturbances culminated in two local blockage experiments in the KNS sodium loop simulating LMFBR fuel elements with a 49% central and a 21% corner blockage. In the frame of this work pin cooling in the wake of the blockage was investigated in single-phase conditions, in boiling conditions up to dryout and in conditions simulating gas release from failed pins. The general aims of the studies were to demonstrate that the consequences of a local blockage do not lead to rapid propagation of damage within a pin bundle and to obtain data for validation of theoretical models.

Little is known about the two-phase pressure loss, the flow pattern, and the critical heat flux conditions for boiling sodium under forced convection. The specific thermohydraulic properties of sodium prohibit extrapolation to sodium of experimental data obtained for other liquids. Therefore, some new test series were carried out in a sodium loop with an induction heated test section of 9 mm inner diameter and 200 mm heated length. The two-phase pressure loss and the film thickness were measured up to the critical cooling conditions. The experimental results are compared with values predicted by known models on annular flow and annular mist flow, respectively. Satisfactory predictions of the flow pattern and the critical heat flux conditions could only be obtained using the measured two-phase pressure losses.

A series of experiments has been carried out using an electrically heated seven pin bundle to simulate the conditions under which boiling of the sodium coolant could occur in the event of the loss of power to the circulator pumps of a fast reactor, coincident with a failure of the reactor to trip. Although it was not possible to represent the conditions of the reactor exactly, nor to continue the tests far into dryout, the results nevertheless give valuable qualitative information on the course of boiling development as well as useful quantitative information against which the predictions of computer codes can be checked. In particular, data have been obtained relating to the incidence of superheat, the location and time to dryout of the residual liquid films, the void fraction within the boiling region, and the types of flow regimes which may be expected within different parts of the boiling region at various stages of the transient.

During research and development work on the SNR-300 sodium-cooled fast reactor the consequences and confinement of sodium fires occurring in enclosures were studied. The behavior of liquid sodium during fires and the behavior of an inherently ready-for-operation protective system are described. Theoretical considerations on the behavior of burning liquid sodium are compared with experimental results. A protective system for large facilities is presented and the use of extinguishing powders is reviewed.

Two series of quasi-steady state sodium boiling experiments have been carried out in an electrically heated seven-pin bundle. The power levels (130–170 and 30–40 W/cm2) and other test conditions were selected to correspond to the core and radial breeder zones of a typical LMFBR. The test procedure involved the gradual reduction of mass flow rate through the bundle in a simulation of the consequences of a slowly growing blockage in the lower part of a reactor subassembly. By this means it was possible to study the development of quasi-steady state boiling up to the onset of permanent dryout. The results obtained provide information on flow regimes in the two-phase region, vapour qualities and flow rates at which cooling of the bundle can be effectively maintained, and the ultimate incidence of dryout. A relation for the two-phase pressure drop multiplier obtained from adiabatic pressure drop measurements in this geometry is given and compared with earlier results from single-channel geometry tests.

A series of sodium boiling experiments, in which the thermohydraulic characteristics of KNK II driver subassemblies were simulated, has been carried out for the purpose of studying the effects of two types of cooling disturbances: rapid flow interruption and those flow reductions which develop rather gradually. Information about the spatial and temporal development of boiling, voidage and dryout was obtained. Furthermore the feasibility of individual subassembly temperature monitoring has been investigated as well as that of two integral boiling detection methods based on acoustic noise and reactivity measurements.

The SIMBATH out-of-pile experiments simulate severe accidents in fast breeder reactors. In the tests the nuclear energy released is substituted by the exothermal energy of a thermite reaction. Single pin and small bundle experiments as well as freezing tests are performed. Material ejected from the fuel rod simulators in an early phase is finely dispersed. A portion penetrates the upper breeding zone without freezing. The bulk of molten material ejected afterwards leads to blockages in the colder zones of the bundle. Under these conditions bottled-up situations may occur in the SIMBATH experiments. Residual sodium may become entrapped. The current version of the computer code CALIPSO developed to interpret these experiments is verified by calculation of two single pin experiments. The computations show that the relocation mechanisms in the SIMBATH experiments are mainly controlled by expansion of noncondensible gases originally existing inside the pins. The contribution from fuel vapour pressure or from sodium evaporation due to fuel-coolant-interaction is of less importance during the first 100 ms after fuel pin failure.

This paper reports on experimental and theoretical investigations of the thermodynamics and hydrodynamics of cooling perturbations in fuel elements of sodium cooled fast reactors. Particular attention is given to bubble nucleation and to growth and shrinkage of the sodium vapour bubble volume. The theoretical model on which the digital computer code BLOW2 is based, is described briefly. Boiling experiments which were carried out in a test loop under typical loss of coolant flow situations of a sodium cooled reactor are analysed with respect to the vapour bubble dynamics, the pressure pulses during the boiling process and the temperatures of the test section wall. The influence of artificial surface cavities and sodium impurities upon superheat is investigated in a separate experiment.