The Best Estimate Plus Uncertainty (BEPU) method is applied to analysis of the “intentional depressurization of steam generator secondary side” which is an accident management procedure in a small break loss-of-coolant accident with high pressure injection system failure. In the present study, scaling calculations from the LSTF small break LOCA tests were carried out for a conventional Westinghouse type four-loop PWR. The two test cases were selected with different break size and different depressurization conditions to ensure the reliability for the analyses for the accident scenario. The uncertainty propagation analyses were performed for the PWR. The dominating input uncertainty parameters, i.e. those few with the highest influence on the output uncertainty were identified. The analysis results were compared with those for the LSTF to address the scaling up capability and the similarity evaluation between the LSTF model and the PWR plant model. It was found that the PWR plant model results had overall agreements with the LSTF model, and the uncertainty of the predicted PCT included the measured PCT. Furthermore, the correlation coefficients between the input uncertainty parameters and the PCT for the PWR plant model had similarities with those for the LSTF model.
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2018 26th International Conference on Nuclear Engineering
July 22–26, 2018
London, England
Conference Sponsors:
- Nuclear Engineering Division
ISBN:
978-0-7918-5149-4
PROCEEDINGS PAPER
Uncertainty Analysis of Scaling Calculations From LSTF Small Break LOCA Tests With Steam Generator Intentional Depressurization Applying to a Four-Loop PWR
Ikuo Kinoshita
Ikuo Kinoshita
Institute of Nuclear Safety System, Inc., Fukui, Japan
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Ikuo Kinoshita
Institute of Nuclear Safety System, Inc., Fukui, Japan
Paper No:
ICONE26-81920, V06BT08A012; 8 pages
Published Online:
October 24, 2018
Citation
Kinoshita, I. "Uncertainty Analysis of Scaling Calculations From LSTF Small Break LOCA Tests With Steam Generator Intentional Depressurization Applying to a Four-Loop PWR." Proceedings of the 2018 26th International Conference on Nuclear Engineering. Volume 6B: Thermal-Hydraulics and Safety Analyses. London, England. July 22–26, 2018. V06BT08A012. ASME. https://doi.org/10.1115/ICONE26-81920
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