Large Scale Verification of External RPV Cooling in Case of Severe Accident
- Framatome ANP GmbH, P.O. Box 3220, D91050 Erlangen (Germany)
The design of the SWR 1000 - developed by Framatome ANP, consists of components, to flood the exterior of the reactor pressure vessel (RPV) in the case of a hypothetical core melt accident. FANP performed tests to demonstrate that there are significant safety margins against occurring of a critical heat flux (CHF). For this purpose different pretests have been performed in a water / air operated test facility. Within the first pretests the global flow conditions around the RPV have been investigated by measuring the local void fractions with impedance and fiber optical probes. In addition the local water velocities have been measured with a Laser Doppler Anemometer. In further pretests a section model has been implemented in the test facility and the geometry has been modified until the flow conditions in the section model and the global model have been similar. Scaling procedures proved, that the water/ air tests of the section model could be transferred to a water /steam operated heated 1:1 scaled model. Such a model has been manufactured and integrated in the BENSON test-rig - a highly flexible water/steam separate effect test-facility operated by Framatome ANP. The model has been equipped with heating wires pressed in slots in the surface of the model, which represented the RPV- wall. The distance between the slots has been chosen in such a way that the decay heat flux profile, which would have to be removed in a core melt accident, could be simulated. In addition more than 300 thermocouples have been installed on the heated surface to measure the wall temperature and observe whether a CHF has occurred. During the tests no CHF occurred - the tests have been limited only by the test set-up. The tests have been performed at ambient conditions, whereas the pressure under accident conditions would be significantly above these conditions. The critical heat flux increases, with increasing pressure. Considering this effect the proven margin against occurring of a CHF is about four. As the measurements have been limited by the test set-up and as the literature shows that CHF-values of inclined and curved comparable structures are much higher, than the tested heat fluxes it can be expected that the real safety margins against occurring of a CHF are much higher than four. (author)
- Research Organization:
- American Nuclear Society, 555 North Kensington Avenue, La Grange Park, IL 60526 (United States)
- OSTI ID:
- 21160696
- Resource Relation:
- Conference: ICAPP'04: 2004 international congress on advances in nuclear power plants, Pittsburgh, PA (United States), 13-17 Jun 2004; Other Information: Country of input: France; 19 refs; Related Information: In: Proceedings of the 2004 international congress on advances in nuclear power plants - ICAPP'04, 2338 pages.
- Country of Publication:
- United States
- Language:
- English
Similar Records
Analysis of two-phase flow phenomena with FLUENT-4 code in the experiments for advanced light water reactor safety
Experimental and Numerical Investigation of Boron Dilution Transients in Pressurized Water Reactors