Good quality experimental data is needed to refine the thermal hydraulic models for the prediction of rod bundle void distribution and critical heat flux (CHF) or dry-out. The Nuclear Power Engineering Corporation (NUPEC) has provided a valuable database to evaluate the thermal hydraulic codes [1]. Part of this database was selected for the NUPEC BWR Full-size Fine-Mesh Bundle Tests (BFBT) benchmark sponsored by US NRC, METI-Japan, NEA/OECD and Nuclear Engineering Program of the Pennsylvania State University (PSU). Twenty-five organizations from ten countries have confirmed their intention to participate and will provide code predictions to be compared to the measured data for a series of defined exercises within the framework of the BFBT benchmark. This benchmark data includes both the fine-mesh high quality sub-channel void fraction and critical power data. Using a full BWR rod bundle test facility, the void distribution was measured at mesh sizes smaller than the sub-channel by using a state-of-the-art computer tomography (CT) technology [1]. Experiments were performed for different pressures, flow rates, exit qualities, inlet sub-cooling, power distributions, spacer types and assembly designs. There are microscopic and sub-channel averaged void fraction data from the CT scanner at the bundle exit as well as X-ray densitometer void distribution data at different elevation levels in the rod bundle. Each sub-channel’s loss coefficient was calculated with using the Rehme method [2,3], and a COBRA-TF sub-channel model was developed for the NUPEC facility. The BWR assembly that was modeled with COBRA-TF includes two water rods at the center. The predicted sub-channel void fraction values from COBRA-TF are compared with the bundle exit void fraction values measured using the CT-scanner void fraction from the BFBT benchmark data. Different plots are used to examine the code prediction of the void distribution at a sub-channel level for the different sub-channels within the bundle.
Skip Nav Destination
14th International Conference on Nuclear Engineering
July 17–20, 2006
Miami, Florida, USA
Conference Sponsors:
- Nuclear Engineering Division
ISBN:
0-7918-4243-6
PROCEEDINGS PAPER
Evaluation of the Sub-Channel Code COBRA-TF for Prediction of BWR Fuel Assembly Void Fraction Distribution Available to Purchase
Fatih Aydogan,
Fatih Aydogan
Pennsylvania State University, University Park, PA
Search for other works by this author on:
Lawrence E. Hochreiter,
Lawrence E. Hochreiter
Pennsylvania State University, University Park, PA
Search for other works by this author on:
Kostadin Ivanov
Kostadin Ivanov
Pennsylvania State University, University Park, PA
Search for other works by this author on:
Fatih Aydogan
Pennsylvania State University, University Park, PA
Lawrence E. Hochreiter
Pennsylvania State University, University Park, PA
Kostadin Ivanov
Pennsylvania State University, University Park, PA
Paper No:
ICONE14-89174, pp. 133-139; 7 pages
Published Online:
September 17, 2008
Citation
Aydogan, F, Hochreiter, LE, & Ivanov, K. "Evaluation of the Sub-Channel Code COBRA-TF for Prediction of BWR Fuel Assembly Void Fraction Distribution." Proceedings of the 14th International Conference on Nuclear Engineering. Volume 2: Thermal Hydraulics. Miami, Florida, USA. July 17–20, 2006. pp. 133-139. ASME. https://doi.org/10.1115/ICONE14-89174
Download citation file:
21
Views
Related Proceedings Papers
Related Articles
Void Fraction Measurement and Prediction of Two-Phase Boiling Flows in a Tubular Test Section
ASME J of Nuclear Rad Sci (April,2023)
Interfacial Heat Transfer Between Steam Bubbles and Subcooled Water in Vertical Upward Flow
J. Heat Transfer (May,1995)
An Analysis of Heat Transfer to Axial Dispersed Flow between Rod Bundles under Reactor Emergency Cooling Conditions
J. Heat Transfer (August,1980)
Related Chapters
Insights and Results of the Shutdown PSA for a German SWR 69 Type Reactor (PSAM-0028)
Proceedings of the Eighth International Conference on Probabilistic Safety Assessment & Management (PSAM)
Completing the Picture
Air Engines: The History, Science, and Reality of the Perfect Engine
Iowa State University Research Reactor
Decommissioning Handbook