Detailed measurements of local parameters in annular Two-Phase FLOW IN FUEL BUNDLE under BWR operating conditions
2015 (English)In: International Topical Meeting on Nuclear Reactor Thermal Hydraulics 2015, NURETH 2015, American Nuclear Society, 2015, 4337-4351 p.Conference paper (Refereed)Text
The Westinghouse FRIGG facility, in Vasteras/Sweden, is dedicated to the measurement of critical power, stability and pressure drop in fuel rod bundles under BWR operating conditions (steady-state and transient). The facility is particularly relevant to test modern BWR fuel designs which typically have complex features, such as part-length rods and mixing vanes that make the flow heterogeneous and challenging to accurately simulate (e.g. using sub-channel analysis codes or CFD tools). In order to support the validation of advanced thermal-hydraulics codes for detailed BWR fuel assembly simulation, new local instrumentation techniques have been tested at the FRIGG facility for the measurement of two-phase dynamic pressure (Pitot tubes) and high time resolution phase detection (optical sensor). The optical sensors were custom-made by RBI Instrumentation for the FRIGG facility and optimized for annular two-phase flow (drop/steam) under BWR operating conditions. This new instrumentation was successfully tested and allows the first-time measurement, under BWR operating conditions, of relevant two-phase flow parameters such as the local void fraction in the steam core, the local drop/steam velocity, the volumetric interfacial area, the drop collision frequency and the assessment of drop size distribution during BWR steady-state and transient operations.
Place, publisher, year, edition, pages
American Nuclear Society, 2015. 4337-4351 p.
Annular two-phase flow, BWR, Instrumentation, Rod bundle, Void fraction, Boiling water reactors, Computational fluid dynamics, Drops, Fuels, Hydraulics, Nuclear reactors, Optical sensors, Parameter estimation, Drop size distribution, High-time resolution, Instrumentation techniques, Rod bundles, Steady state and transients, Thermal hydraulics codes, Two phase flow
IdentifiersURN: urn:nbn:se:kth:diva-187510ScopusID: 2-s2.0-84962632728ISBN: 9781510811843OAI: oai:DiVA.org:kth-187510DiVA: diva2:937666
16th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2015, 30 August 2015 through 4 September 2015
QC 201606152016-06-152016-05-252016-06-15Bibliographically approved