CHF prediction in nuclear fuel elements by using round tube data |
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| Affiliation: | 1. Universidade Federal de Minas Gerais, Departamento de Engenharia Nuclear, Av. Contorno 842, 30110-060 Belo Horizonte, MG, Brazil;2. Comissão Nacional de Energia Nuclear, Centro de Desenvolvimento da Tecnologia Nuclear, Rua Prof. Mario Werneck s/n, Cidade Universitária, 30123-970 Belo Horizonte, MG, Brazil;1. Heilongjiang Provincial Key Laboratory of Nuclear Power System & Equipment, Harbin Engineering University, Harbin 150001, China;2. China Nuclear Power Engineering Co. Ltd, Beijing 100840, China;1. School of Nuclear Science and Technology, Shaanxi Key Laboratory of Advanced Nuclear Energy and Technology, State Key Laboratory of Multiphase Flow in Power Engineering, Xi''an Jiaotong University, Xi''an 710049, China;2. Shanghai Nuclear Engineering Research & Design Institute, Shanghai 200233, China;1. Los Alamos National Laboratory, P.O. Box 1663, Los Alamos, NM 87545, United States;2. Design Physics Department, AWE Aldermaston, Berkshire RG7 4PR, UK |
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| Abstract: | The 1995 CHF table for uniformly heated round tubes, developed jointly by Canadian and Russian researchers, has been used for the prediction of critical heat flux (CHF) in 5×5 test sections simulating fuel elements of pressurized water reactors. Comparisons between measured and calculated CHF indicates that the table with an appropriate diameter correction can be applied to rod bundles of the type considered in this study. The relation for the diameter correction factor was derived from the CHF data. The tolerance limits associated with the departure from nucleate boiling ratio (DNBR) are evaluated by using statistical analysis. |
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