A sensitivity study of physical models using in RELAP5 code based on FEBA experimental data
Main Article Content
In the thermal-hydraulic safety analysis, simulation results using thermal-hydraulic codes depend mainly on modeling the physical phenomena built-in the codes. These models are the equations, and empirical formulas developed based on matching them to experimental data or based on the assumptions, simplifications for solving theoretical equations. Therefore, it is recommended that these physical models need to take into account the uncertainty they cause. The sensitivity study is performed to investigate the influence of physical models on the calculation results during the reflood phase after the loss of coolant accident. It is allowable to choose the physical models that have the most significant influence on the calculation results. This study conducted a sensitivity analysis of physical models in RELAP5 code based on experimental data measured on the FEBA test facility. Sixteen physical models have been selected for sensitivity analysis to find the most important models that influence the calculation results. Based on two criteria, the maximum cladding temperature and the quench time, the sensitivity analysis results show that four physical models significantly impact the calculation result. Four chosen physical models are considered further in the next step of their uncertainty evaluation.
physical model, FEBA, sensitivity, uncertainty, quench time, PCT
. ISL, RELAP5/MOD3.3 code manual volume IV: models and correlations, NUREG/CR5535/Rev P3-Vol IV, 2006.
. E Elias, Rewetting and liquid entrainment during reflooding - state of the art, EPRI NP435, (Research Project 248-1), Topical Report, May 1977.
. NEA, Nuclear fuel behaviour in loss-of-coolant accident (LOCA) conditions: State-of-the-art Report, Nuclear Energy Agency, 2009.
. Choi T. S., No H. C., Improvement of the reflood model of RELAP5/MOD3.3 based on the assessments against FLECHT-SEASET tests, Nuclear Engineering and Design, Vol. 240, pp.832–841, 2010.
. Cadek F.F., Dominics D. P., Layse R. H., PWR FLECHT Final Report, WCAP-7665, 1971.
. Lee N. et al., PWR FLECHT-SEASET unblocked bundle, forced and gravity relood task data evaluation and analysis report, NUREG/CR-2256, 1982.
. G.H. Seo et al. Numerical analysis of RBHT reflood experiments using MARS 1D and 3D modules, Journal of Nuclear Science and Technology, Vol. 52, pp.70-84, 2015.
. Hochreiter L.E et al., RBHT relood heat transfer experiments data and analysis, NUREG/CR-6980, 2012.
. P. Ihle, K. Rust, FEBA Flooding Experiments with Blocked Arrays Evaluation Report, März 1984.
. A. Kovtonyuk et al., Post-BEMUSE Reflood Model Input Uncertainty Methods (PREMIUM) Benchmark Phase II: Identification of Influential Parameters, NEA/CSNI/R(2014)14, 2015.
. M. Perez et al., Uncertainty and sensitivity analysis of a LBLOCA in a PWR Nuclear Power Plant: Results of the Phase V of the BEMUSE programme, Nuclear Engineering and Design, Vol. 241, pp. 4206 – 4222, 2011.
. Horst Glaeser, GRS Method for Uncertainty and Sensitivity Evaluation of Code Results and Applications, Science and Technology of Nuclear Installations, pp. 1-7, 2008.
. Sencar M., and Aksan N., Evaluation and Assessment of Reflooding Models in RELAP5/MOD2.5 and RELAP5/MOD3 Codes Using Lehigh University and PSINEPTUN Bundle Experimental Data, Seventh International Meeting on Nuclear Reactor Thermohydraulics (NURETH-7), Saratoga Springs, NY, pp. 2280–2302, 1995.
. J. Bánáti, Analysis of REWET-II Reflooding Experiments with RELAP5/MOD3, 1994.
. T.S. Choi and HC NO, An improved RELAP5/MOD3.3 reflood model considering the effect of spacer grids, Nuclear Engineering and Design, Vol. 250, pp. 613–625, 2012.
. D. Li et al., improvement of reflood model in RELAP5 code based on sensitivity analysis, Nuclear Engineering and Design, Vol. 303, pp.163–172, 2016.
. C. Berna et al., Review of droplet entrainment in annular flow: Characterization of the entrained droplets, Progress in Nuclear Energy, Vol. 79, pp. 64-86, 2015.
. J.C. Chen, Correlation for Boiling Heat Transfer to Saturated Fluids in Convective Flow, Industrial & Engineering Chemistry Process Design and Development, Vol. 3, pp. 322-329, 1966.
. H. K. Forster N. Zuber, Dynamics of vapor bubbles and boiling heat transfer, A.I.Ch.E. Journal, Vol. 1, pp. 521-535, 1955.
. F. W. Dittus, L. M. K. Boelter, Heat transfer in automobile radiators of the tubular type, The University of California Publications on Engineering, Vol.2, pp. 443-461, 1930.
. R. H. S., Winterton, Where did the Dittus and Boelter equation come from?. Int. J. Heat Mass Tran., Vol. 41, pp.809-810, 1998.
. W. H. McAdams, Heat Transmission, 3rd edition, McGraw-Hill, New York, 1954.
. D.C. Groeneveld et al., The 2006 CHF look-up table, Nuclear Engineering and Design, Vol. 237, pp.1909–1922, 2007.
. Y. Taitel, D. Bornea, and A. E. Dukler. “Modeling Flow Pattern Transitions for Steady Upward Gas-Liquid Flow in Vertical Tubes”. AIChE Journal, Vol. 26, pp. 345-354, 1980.
. J. G. M. Andersen and H. Abel-Larsen, CORECOOL-Model Description of the Program, RISO-M-21380, November 1980.