No abstract available

[en] An unintentional reduction of the coolant flow affects the energy output from the core, and consequently the safety of the reactor. Possible causes of a reduction of the coolant flow are presented. Possibilities of reducing the required NPSH-values for the pumps and the necessity for testing are discussed, followed by a description of the pump test bed at the GKSS. In an abstract on special measuring methods the development of a testing procedure for the indication of cavitation in coolant pumps, and of a measuring device for the measurement of partial gas pressures, and the available NPSH-value at the pump inlet nozzle respectively are described

No abstract available

[en] Short communication; 3 refs.; and, 2 figures

[en] A recommended position for the thermocouples used in the RSG-GAS loss of flow transient, as documented in the RSG-GAS experimental description report, is provided in the following table and figure. The recommended thermocouple positions were agreed among the CRP participants and consolidated during the Consultancy Meeting held at IAEA headquarters, Vienna, 15–18 July 2013.

No abstract available

[en] Two unprotected loss-of-heat sink transients initiated from near full power conditions in the Experimental Breeder Reactor-II (EBR-II) plant have been simulated. In one transient the secondary sodium flow is reduced to nearly zero (0.5% of its initial value) while in the other the flow simply coasts down to a natural-convective rate of about 8%. In spite of the large difference in primary heat removal rates, which the difference in secondary flow rates represents, both transients have very similar overall behavior. In addition, the large volume of sodium in the primary tank causes a slowly rising tank temperature in response to net heat addition. An important result is that the negative reactivity feedback characteristics of the reactor cause it to shut itself down in a benign manner in both cases. Experiments based on these simulations are planned for the EBR-II in 1985

[en] Short communication

[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

[en] Two experiments, each using an hexagonal array of 37 full-length mixed-oxide fuel pins, have simulated an unprotected loss-of-flow accident under prototypic sodium-cooled fast-reactor conditions. These two experiments are summarized and experimental results are compared with accident analysis calculations. It is shown that the voiding behavior predicted by the SAS3D accident analysis code closely agrees with the measurements but the rates of blockage formation and fuel melting are somewhat over estimated by the code. 8 refs