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Thermal-hydraulic behavior simulations of the reactor cavity cooling system (RCCS) experimental facility using Flownex

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초록

The scaled water-cooled Reactor Cavity Cooling System (RCCS) experimental facility reproduces a passive safety feature to be implemented in Generation IV nuclear reactors. It keeps the reactor cavity and other internal structures in operational conditions by removing heat leakage from the reactor pressure vessel. The present work uses Flownex one-dimensional thermal-fluid code to model the facility and predict the experimental thermal-hydraulic behavior. Two representative steady-state cases defined by the bulk volumetric flow rate are simulated (Re ¼ 2,409 and Re ¼ 11,524). Results of the cavity outlet temperature, risers' temperature profile, and volumetric flow split in the cooling panel are also compared with the experimental data and RELAP system code simulations. The comparisons are in reasonable agreement with the previous studies, demonstrating the ability of Flownex to simulate the RCCS behavior. It is found that the low Re case of 2,409, temperature and flow split are evenly distributed across the risers. On the contrary, there's an asymmetry trend in both temperature and flow split distributions for the high Re case of 11,524.

Abstract

The scaled water-cooled Reactor Cavity Cooling System (RCCS) experimental facility reproduces a passive safety feature to be implemented in Generation IV nuclear reactors. It keeps the reactor cavity and other internal structures in operational conditions by removing heat leakage from the reactor pressure vessel. The present work uses Flownex one-dimensional thermal-fluid code to model the facility and predict the experimental thermal-hydraulic behavior. Two representative steady-state cases defined by the bulk volumetric flow rate are simulated (Re ¼ 2,409 and Re ¼ 11,524). Results of the cavity outlet temperature, risers' temperature profile, and volumetric flow split in the cooling panel are also compared with the experimental data and RELAP system code simulations. The comparisons are in reasonable agreement with the previous studies, demonstrating the ability of Flownex to simulate the RCCS behavior. It is found that the low Re case of 2,409, temperature and flow split are evenly distributed across the risers. On the contrary, there's an asymmetry trend in both temperature and flow split distributions for the high Re case of 11,524.

참고문헌 (18)

  1. World Energy Outlook 2021, "IEA", IEA, ~ (2021)
  2. IAEA, "Energy, Electricity and Nuclear Power Estimates for the Period up to, 2050" (2020), IAEA
  3. G. Locatelli, "Generation IV nuclear reactors : current status and future prospects", Energy Pol, 61, pp.1503~1520 (2013)
  4. J. Thielman, "Evaluation and optimization of general atomics' gtmhr reactor cavity cooling system using an axiomatic design approach", Nucl. Eng. Des, 235(13), pp.1389~1402 (2005)
  5. L. J. Lommers, "AREVA HTR concept for near-term deployment", Nucl. Eng. Des, 251, pp.292~296 (2012)
  6. R. Vaghetto, "Experimental investigation of a scaled water-cooled reactor cavity cooling system", Nucl. Technol, 187(3), pp.282~293 (2014)
  7. R. Vaghetto, "Modeling the thermalehydraulic behavior of the reactor cavity cooling system using RELAP5-3D", Ann. Nucl. Energy, 73, pp.75~83 (2014)
  8. N. R. Quintanar, "Natural circulation flow distribution within a multi-branch manifold", Int. J. Heat Mass Tran, 153, pp.1~15 (2019)
  9. D. Holler, "Evaluation and quantification of optic fiber distributed sensor response in air and water environments", Int. J. Therm. Sci, 145, pp.106042~ (2019)
  10. M. Gorman, "Reactor Cavity Cooling System Computational Fluid Dynamic Analysis", Transactions Of The American Nuclear Society (2021)
  11. M. Pehlivan, "Simulations of the Scaled Reactor Cavity Cooling System Experimental Facility with RELAP5", SCDAPSIM System Code, 2021, ~
  12. RELAP5-3D Code Manuals, "R. Schultz", ~ (2005)
  13. "Flownex SE, Version 8.14.0" (2022)[URL]
  14. Flownex General User Manual, "M.- Tech Industrial", M-Tech Industrial, ~ (2022)
  15. P. Rousseau, "Code-to-code comparison for analysing the steady-state heat transfer and natural circulation in an air-cooled RCCS using GAMMAþ and Flownex", Nucl. Eng. Des, 291, pp.71~89 (2015)
  16. C. G. du Toit, "Fundamental evaluation of the effect of pipe diameter, loop length and local losses on steady-state single-phase natural circulation in square loops using the 1D network code Flownex", Therm. Sci. Eng. Prog, 22 (2021)
  17. G. P. Greyvenstein, "An implicit method for the analysis of transient flows in pipe networks", Int. J. Numer. Methods Eng, 53(5), pp.1127~1143 (2002)
  18. Flownex Theory Manual, "M. -Tech Industrial", M-Tech Industrial, ~ (2022)

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