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Uncertainty analysis of heat transfer of TMSR-SF0 simulator

Dai Ye|Wang Jiajun|Zou Yang|Xu Hongjie
한국원자력학회 (Korean Nuclear Society)|2024.02|Nuclear Engineering and Technology|제56권 제2호|pp. 762-769|DOI 10.1016/j.net.2023.11.016

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Heat transferUncertainty analysisThermal-hydraulicTMSR-SF0
주제분류원자력공학

초록

The TMSR-SF0 simulator is an integral effect thermal-hydraulic experimental system for the development of thorium molten salt reactor (TMSR) program in China. The simulator has two heat transport loops with liquid FLiNaK. In literature, the 95% level confidence uncertainties of the thermophysical properties of FLiNaK are recommended, and the uncertainties of density, heat capacity, thermal conductivity and viscosity are ±2%, ±10, ±10% and ±10% respectively. In order to investigate the effects of thermophysical properties uncertainties on the molten salt heat transport system, the uncertainty and sensitivity analysis of the heat transfer characteristics of the simulator system are carried out on a RELAP5 model. The uncertainties of thermophysical properties are incorporated in simulation model and the Monte Carlo sampling method is used to propagate the input uncertainties through the model. The simulation results indicate that the uncertainty propagated to core outlet temperature is about ±10 ◦C with a confidence level of 95% in a steady-state operation condition. The result should be noted in the design, operation and code validation of molten salt reactor. In addition, more experimental data is necessary for quantifying the uncertainty of thermophysical properties of molten salts.

Abstract

The TMSR-SF0 simulator is an integral effect thermal-hydraulic experimental system for the development of thorium molten salt reactor (TMSR) program in China. The simulator has two heat transport loops with liquid FLiNaK. In literature, the 95% level confidence uncertainties of the thermophysical properties of FLiNaK are recommended, and the uncertainties of density, heat capacity, thermal conductivity and viscosity are ±2%, ±10, ±10% and ±10% respectively. In order to investigate the effects of thermophysical properties uncertainties on the molten salt heat transport system, the uncertainty and sensitivity analysis of the heat transfer characteristics of the simulator system are carried out on a RELAP5 model. The uncertainties of thermophysical properties are incorporated in simulation model and the Monte Carlo sampling method is used to propagate the input uncertainties through the model. The simulation results indicate that the uncertainty propagated to core outlet temperature is about ±10 ◦C with a confidence level of 95% in a steady-state operation condition. The result should be noted in the design, operation and code validation of molten salt reactor. In addition, more experimental data is necessary for quantifying the uncertainty of thermophysical properties of molten salts.

참고문헌 (49)

  1. An overview of liquid-fluoride-salt heat transport systems, "D. E. Holcomb", 990239~, pp.ORNL (2010), TM-2010[DOI]
  2. Operation of the aircraft reactor experiment, Tech. rep, "W. Cottrell", Oak Ridge National Lab, ~ (1955)
  3. MSRE design & operations report part 1 description of reactor design, Tech. rep, "R. C. Robertson", Oak Ridge National Lab, ~ (1965)
  4. Conceptual design study of a single-fluid molten-salt breeder reactor, Tech. rep, "R. C. Robertson", Oak Ridge National Lab, ~ (1971)
  5. M. D. Grele, "Forced-Convection Heat-Transfer Characteristics of Molten FLiNaK Flowing in an Inconel X System", National Advisory Committee for Aeronautics (1954)
  6. A. Bergman, "General phase diagram for KF–LiF–NaF system", Compt. Rend. Acad. Sci, 31, pp.753~755 (1941)
  7. Fused salt heat transfer. Part II. Forced convection heat transfer in circular tubes containing NaF-KF-LiF eutectic, Tech. rep, "H. Hoffman", Oak Ridge National Lab, ~ (1955)
  8. Molten salt heat transfer, Tech. rep, "H. Hoffman", Oak Ridge National Lab, ~ (1958)
  9. Forced-convection heat-transfer measurements with a molten fluoride salt mixture flowing in a smooth tube, Tech. rep, "J. Cooke", Oak Ridge National Lab, ~ (1973)
  10. V. Ignat’Ev, "Heat exchange during the flow of a melt of lif-naf-KF fluoride salts in a circular tube", Soviet Atomic Energy, 57(2), pp.560~562 (1984)
  11. Heat transfer measurements in a forced convection loop with two molten-fluoride salts: LiF–BeF 2–ThF 2–UF 4 and eutectic NaBF 4–NaF, Tech. rep, "M. Silverman", Oak Ridge National Lab, ~ (1976)
  12. Liquid fluoride salt experiment using a small natural circulation cell, "G. Yoder", Oak Ridge National Laboratory, ~, pp.ORNL (2014), TM-2014[DOI]
  13. B. Vriesema, "Aspects of molten fluorides as heat transfer agents for power generation", ~ (1979)
  14. Assessment of properties of candidate liquid salt coolants for the advanced high temperature reactor (AHTR), "D. Williams", 1~38, pp.ORNL (2005), GEN4[DOI]
  15. J. Ambrosek, "Current status of knowledge of the fluoride salt(FLiNaK)heat transfer", Nucl. Technol, 165(2), pp.166~173 (2009)
  16. R. C. Gallagher, "Assessment of molten eutectic LiF-NaF-KF density through experimental determination and semiempirical modeling", J. Chem. Eng. Data, 67(6), pp.1406~1414 (2022)
  17. Assessment of Candidate Molten Salt Coolants for the Advanced High Temperature Reactor(AHTR), "D. F. Williams", United States. Department of Energy, ~ (2006)
  18. R. Romatoski, "Fluoride salt coolant properties for nuclear reactor appli-cations: A review", Ann. Nucl. Energy, 109, pp.635~647 (2017)[DOI]
  19. Fused salt heat transfer Part II: forced convection heat transfer in circular tubes containing NaF-KF-LiF entectic, Tech. rep, "H.W. Hoffman", ORNL, ~, pp.ORNL-1777 (1955)
  20. G. Yoder, "Examination of liquid fluoride salt heat transfer", Proceedings of ICAPP, pp.6~9 (2014)
  21. E. de Rocquigny, "Uncertainty in Industrial Practice: A Guide To Quantitative Uncertainty Management" (2008), John Wiley & Sons
  22. M. Panicker, "Licensing and regulatory requirements for best estimate plus uncertainty applications", Proc. Int. Conf. Best Estimate Plus Uncertainty (BEPU 2018), pp.13~18 (2018)
  23. C. Wang, "Thermal-hydraulic design and analysis of a small modular molten salt reactor(MSR)with solid fuel", Int. J. Energy Res, 42(3), pp.1098~1114 (2018)
  24. Y. Zhao, "The uncertainty analysis of natural circulation in molten salt reactor coupling dakota with GenFlow", International Conference on Nuclear Engineering, 57847, pp.V006T08A068~ (2017), American Society of Mechanical Engineers
  25. Uncertainty quantification approaches for advanced reactor analyses, "L. Briggs", ~ (2009)
  26. R. R. Romatoski, "Fluoride-salt-cooled high-temperature test reactor thermal-hydraulic licensing and uncertainty propagation analysis", Nucl. Technol, 205(11), pp.1495~1512 (2019)
  27. J. Serp, "The molten salt reactor(MSR)in generation IV : Overview and perspectives", Prog. Nucl. Energy, 77, pp.308~319 (2014)[DOI]
  28. D. Jiang, "Fluoride-salt-cooled high-temperature reactors: Review of historical milestones, research status, challenges, and outlook", Renew. Sustain. Energy Rev, 161, pp.112345~ (2022)[DOI]
  29. M. Jiang, "Advanced fission energy program-TMSR nuclear energy system", Bulletin of Chinese Academy of Sciences, 27, pp.366~374 (2012)
  30. Z. Dai, "Molten Salt Reactors and Thorium Energy", pp.531~540 (2017), Elsevier
  31. T. J. Dolan, "Woodhead Publishing Series in Energy" (2017), Woodhead Publishing
  32. "Analysis on ucrw-atws in tmsr-sf1" (2015)[URL]
  33. J. Chen, "Coupled neutronic and thermal-hydraulic analysis of TMSR-SF1 at steady state", Nucl. Tech, 40, pp.83~90 (2017)
  34. Y. Liu, "Sensitivity", uncertainty comparison and similarity analysis between TMSR-LF1 and MSR models, 2020, pp.122, 103289~
  35. H. Xu, "Thorium molten salt reactor and comprehensive utilization of nuclear energy", Mod. Phys (2018)[DOI]
  36. G. Zheng, "Thorium-based molten salt SMR as the nuclear technology pathway from a market-oriented perspective", Ann. Nucl. Energy, 116, pp.177~186 (2018)[DOI]
  37. J. Wang, "Transient analysis of TMSR-SF0 simulator", J. Nucl. Sci. Technol, 58(3), pp.292~301 (2021)[DOI]
  38. W. Liu, "Digital Twin-Driven Development of Online Monitoring and Data Management Systems in TMSR-SF0", International Conference on Nuclear Engineering, 12 (2022), Innovative and Smart Nuclear Power Plant Design
  39. G. Huang, "Nuclear Power Plants: Innovative Technologies for Instrumentation and Control Systems", pp.112~119 (2020), Springer Singapore
  40. Relap5, "mod3.3 code manual volumn i:code structure, system models, and solution methods, Tech. rep", Information System Laboratories, Inc, 2001
  41. X. Jiao, "Core safety discussion under station blackout ATWS accident of solid fuel molten salt reactor", Nucl. Tech, 38 (2015)
  42. S. Jiang, "Internal Topical Meeting on Nuclear Reactor Thermal Hydraulics" (2015)
  43. C. Shi, "Development and application of a system analysis code for liquid fueled molten salt reactors based on RELAP5 code", Nucl. Eng. Des, 305, pp.378~388 (2016)[DOI]
  44. R. Li, "Improvement of the delayed neutron precursor transport model in RELAP5 for liquid-fueled molten salt reactor", Nucl. Eng. Des, 394, pp.111817~ (2022)[DOI]
  45. B. Xu, "Accident analyses of station blackout for TMSR-SF2", Nucl. Tech, 40 (2017)
  46. T. Xu, "ATWS accident analysis of rod withdrawal in small modular molten salt reactor", Nucl. Tech, 45 (2022)
  47. R.R. Romatoski, "Fluoride-Salt-Cooled High-Temperature Test Reactor ThermalHydraulic Licensing and Uncertainty Propagation Analysis", Ph.D, pp.2017
  48. S. S. Wilks, "Determination of sample sizes for setting tolerance limits", Ann. Math. Stat, 12(1), pp.91~96 (1941)
  49. N. Porter, "Wilks’ formula applied to computational tools : A practical discussion and verification", Ann. Nucl. Energy, 133, pp.129~137 (2019)[DOI]

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