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| Book/Report | FZJ-2018-06967 |
1994
Forschungszentrum Jülich GmbH Zentralbibliothek, Verlag
Jülich
Please use a persistent id in citations: http://hdl.handle.net/2128/20272
Report No.: Juel-2961
Abstract: For the analysis of leaks and transients in Power reactors, the thermohydraulics code ATHLET has been developed at the GRS. In order to extend the application area of this code to the safety analysis of research reactors, a rnodel is implemented to describe the thermodynarnic nonequilibriurn effects in subcooled boiling regime. The aim is to simulate void distribution and thermodynamric instability, which is practicularly pronounced in research reactors due to high power densities and low system pressures, and to include the influence of the steam formed in this boiling regime on the neutron balance. The model developed as part of this work considers the competing effects of vaporization and condensation during subcooled boiling. lt describes the rate of bubble generation on superheated surfaces and the subsequent condensation of steam in the subcooled liquid. The installed model is validated by postealculations of two extensive series of experiments. In the first series, the McMaster experiments on axial void distribution in the subcooled boiling regime are recalculated. The postcalculations show that the extented program is capable of caculating the axial void distribution in the subcooled boiling regime with a maximum deviation of approx. 23% from the experiment. The second seriesconcern KFA experiment on thermohydraulic instability during subcooled boiling, which cover a wide parameter range of heat flux density, inlet temperature and channel width. The postcalculations of this experimental series show that the simulation of thermohydraulic instability is ensured by the program extension. The point of beginning instability is conservatively determined in 20 of a total of 27 recaleulated experiments. The extended and verified program is finally used to simulate the Jülich research reactor FRJ-2. For this purpose, a full-scale simulation model of the entire plant is developed ensuring, in particular, a precise reproduction of the geometry and the arrangement of the annular fuel element cooling channels. The modelled reactor plant is first used to simulate normal reactor operation. The resulting steady-state temperature and pressure distributions assuming a thermal power of 23 MW show good agreement with real operating data. Finally, safety investigations are conducted to examine plant behaviour under designbasis accident conditions. This includes failure of all three rnain coolent pumps with proper and delayed reactor scram. In both cases, the simulation shows that the fuel elements are not endangered in any phase of the transient, although in the event of adelayed scram initial signs of parallel channel instability due to steam formation in the central fuel element are to be observed which, however, only prevails for a skort period of 30 ms.
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