In this paper, we deal with a typical pressurizer surge line in a conventional pressurized water reactor (PWR). This study is performed to develop an understanding of thermal stratification phenomenon, which may occur in the surge line during either normal condition or transient process, in the pressurizer surge line. The pressurizer surge line model of Daya Bay nuclear power plant is used as base analysis model, in which the hot leg is taken into account. The transient temperature distribution required to assess the phenomenon along the pressurizer surge line is obtained through CFD analysis technology using ANSYS FLUENT. The temperature loads are transferred to ANSYS Mechanical for stress evaluation for the heat up transient process. Subsequently, the usage factor is calculated on the basis of ASME Section-III design curve. The possible mitigation scheme for the thermal stratification phenomenon of changing the layout angles is also simulated and analyzed in detail. The results show that the thermal stratification phenomenon will occur both in normal operating condition and in heat up transient process. The circumfluent effect makes the thermal stratification phenomenon exhibit unique profile due to the introduction of the hot leg. The continuous spray mass flow rate may influence both the temperature difference and the occurrence range for the thermal stratification phenomenon. The stress analysis incorporating both temperature load and pressure load is performed for pressurizer surge line model with hot leg for the conservative and complete heat up case.
- Nuclear Engineering Division
The Surge Line Stress Analysis Model Setup With the Consideration of Thermal Stratification
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Wang, J, Ge, Z, Sun, Z, & Yan, C. "The Surge Line Stress Analysis Model Setup With the Consideration of Thermal Stratification." Proceedings of the 2014 22nd International Conference on Nuclear Engineering. Volume 4: Radiation Protection and Nuclear Technology Applications; Fuel Cycle, Radioactive Waste Management and Decommissioning; Computational Fluid Dynamics (CFD) and Coupled Codes; Reactor Physics and Transport Theory. Prague, Czech Republic. July 7–11, 2014. V004T10A029. ASME. https://doi.org/10.1115/ICONE22-30686
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