Citation: | Yin Yuan, Feng Simin, Pang Bo, Xi Yanyan, Zhang Yuxiang, Fu Xiangang. Development of Dimensionless Rod-bundle CHF Correlation Based on Stepwise Regression and Determination of DNBR Limit[J]. Nuclear Power Engineering, 2023, 44(4): 72-78. doi: 10.13832/j.jnpe.2023.04.0072 |
[1] |
CHENG X, MÜLLER U. Review on critical heat flux in water cooled reactors: FZKA-6825[R]. Karlsruhe: Forschungszentrum Karlsruhe, 2003.
|
[2] |
张玉相,席炎炎,庞铮铮,等. CHF关系式开发与DNBR限值确定方法研究[J]. 核动力工程,2016, 37(5): 130-134.
|
[3] |
TONG L S, WEISMAN J. Thermal analysis of pressurized water reactors[M]. 3rd ed. La Grange Park: American Nuclear Society, 1996: 478-490.
|
[4] |
REDDY D G, FIGHETTI C F. Parametric study of CHF data Volume 2. A generalized subchannel CHF correlation for PWR and BWR fuel assemblies. Final report: EPRI-NP-2609-Vol. 2[R]. New York: Columbia University, 1983.
|
[5] |
刘伟. 压水堆燃料组件临界热流密度关系式的开发、评估及应用[D]. 西安: 西安交通大学, 2013.
|
[6] |
FIGHETTI C F, REDDY D G. Parametric study of CHF data. Volume 3, Part 1. Critical heat flux data. Final report: EPRI-NP-2609-Vol. 3-Pt. 1[R]. New York: Columbia University, 1982.
|
[7] |
STEWART C W, WHEELER C L, CENA R J, et al. COBRA-IV: the model and the method: BNWL-2214[R]. Richland: Pacific Northwest Laboratories, 1977.
|
[8] |
DRAPER N R, SMITH H. Applied regression analysis[M]. New York: John Wiley & Sons Inc, 1981.
|
[9] |
WEISMAN J, PEI B S. Prediction of critical heat flux in flow boiling at low qualities[J]. International Journal of Heat and Mass Transfer, 1983, 26(10): 1463-1477. doi: 10.1016/S0017-9310(83)80047-7
|
[10] |
LEE C H, MUDAWWAR I. A mechanistic critical heat flux model for subcooled flow boiling based on local bulk flow conditions[J]. International Journal of Multiphase Flow, 1988, 14(6): 711-728. doi: 10.1016/0301-9322(88)90070-5
|
[11] |
TONG L S, TANG Y S. Boiling heat transfer and two-phase flow[M]. 2nd ed. Washington: Taylor & Francis Ltd. , 1997: 333.
|
[12] |
GROENEVELD D C, CHENG S C, DOAN T. 1986 AECL-UO critical heat flux lookup table[J]. Heat Transfer Engineering, 1986, 7(1-2): 46-62. doi: 10.1080/01457638608939644
|
[13] |
ROSAL E R, CERMAK J O, TONG L S, et al. High pressure rod bundle DNB data with axially non-uniform heat flux[J]. Nuclear Engineering and Design, 1974, 31(1): 1-20. doi: 10.1016/0029-5493(74)90129-0
|
[14] |
刘伟, 杜思佳, 张渝, 等. 一种基于分组法的CHF关系式DNBR限值统计学确定方法: 中国, 201910887222.2[P]. 2020-01-24.
|
[15] |
国家质量技术监督局. 数据的统计处理和解释正态性检验: GB/T 4882-2001[S]. 北京: 中国标准出版社, 2004:14-17.
|
[16] |
OWEN D B. Factors for one-sided tolerance limits and for variables sampling plans: NSA-17-023849[R]. Albuquerque: Sandia Corporation, 1963.
|
[17] |
NATRELLA M G. Experimental statistics[M]. Washington: NBS Handbook 91 National Bureau Standards, 1963: 31.
|
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