본문 바로가기
KERD
논문KCI 등재

Experimental Investigation on Critical Heat Flux in Bilaterally Heated Annulus with equal heat flux on both sides

2

조회

0

다운로드

0

인용

0

참고문헌

키워드

주제분류원자력공학

초록

A phenomenological study on CHF in a bilaterally heated annulus with equal heat flux on both sides was experimentally performed. The working fluid of the present test was R-134a. Variation characteristics of CHF and transition of CHF occurrence location were investigated under different pressure, mass flux and quality conditions. With the increase of critical thermodynamic quality, it was found that CHF first occurred on the outer surface of the annulus, then simultaneously occurred on both sides, and finally occurred on the inner surface at relatively high critical quality. After the CHF location transitioned to the inner rod, the sharp fall of CHF in the limiting critical quality region was observed. The critical quality corresponding to the CHF location transition decreased with the increase of mass flux and pressure. Besides, CHF in tube, internally heated, externally heated and bilaterally heated annuli were compared under the same hydraulic diameter conditions. The present study is conducive to improving the understanding of complicated CHF mechanism in bilaterally heated annulus, enriching the experimental database, and providing evidence for developing accurate CHF mechanism model for annuli.

Abstract

A phenomenological study on CHF in a bilaterally heated annulus with equal heat flux on both sides was experimentally performed. The working fluid of the present test was R-134a. Variation characteristics of CHF and transition of CHF occurrence location were investigated under different pressure, mass flux and quality conditions. With the increase of critical thermodynamic quality, it was found that CHF first occurred on the outer surface of the annulus, then simultaneously occurred on both sides, and finally occurred on the inner surface at relatively high critical quality. After the CHF location transitioned to the inner rod, the sharp fall of CHF in the limiting critical quality region was observed. The critical quality corresponding to the CHF location transition decreased with the increase of mass flux and pressure. Besides, CHF in tube, internally heated, externally heated and bilaterally heated annuli were compared under the same hydraulic diameter conditions. The present study is conducive to improving the understanding of complicated CHF mechanism in bilaterally heated annulus, enriching the experimental database, and providing evidence for developing accurate CHF mechanism model for annuli.

참고문헌 (17)

  1. Junliang Guo, "Fluid-to-Fluid modelling of CHF at High-Pressure subcooled water conditions", Nuclear Engineering and Design, 386, pp.111577~ (2022), Elsevier BV[DOI]
  2. Simple but Accurate Round Tube, Uniform Heat Flux, Dryout Correlation over the Pressure Range 0.7 to 17 MN, "m2 (100 to 2500 Psia)", R.W. Bowring, 1972, AEEW-R-789
  3. David D. Hall, "Critical heat flux (CHF) for water flow in tubes—II.", International Journal of Heat and Mass Transfer, 43(14), pp.2605~2640 (2000), Elsevier BV[DOI]
  4. Y. Katto, "An improved version of the generalized correlation of critical heat flux for the forced convective boiling in uniformly heated vertical tubes", Int. J. Heat Mass Tran, 27, pp.1641~1648 (1984)
  5. S. Doerffer, "A comparison of critical heat flux in tubes and annuli", Nuclear Engineering and Design, 149(1-3), pp.167~175 (1994), Elsevier BV[DOI]
  6. K. M. Becker, "Measurements of Burnout Conditions for Flow of Boiling Water in a Vertical Annulus", Journal of Heat Transfer, 86(3), pp.393~404 (1964), ASME International[DOI]
  7. A. P. Ornatskiy, "Burnout in annular channels with two-sided heating", Heat Tran. Sov. Res, 1, pp.12~16 (1969)
  8. V. I. Tolubinskiy, "Heat transfer crisis at the inward surface of annuli heated from both sides", Heat Tran. Sov. Res, 5, pp.93~96 (1973)
  9. S. Doerffer, "A comparison of critical heat flux in tubes and annuli", Nuclear Engineering and Design, 149(1-3), pp.167~175 (1994), Elsevier BV[DOI]
  10. Y.W. Wu, "Experimental study on critical heat flux in bilaterally heated narrow annuli", International Journal of Multiphase Flow, 35(11), pp.977~986 (2009), Elsevier BV[DOI]
  11. Yang Liu, "Critical heat flux measurement and visualization in R-134a on a vertical single rod with and without mixing-vanes spacer: The characteristics and phenomenon of critical heat flux", Annals of Nuclear Energy, 160, pp.108399~ (2021), Elsevier BV[DOI]
  12. Y. Katto, "A generalized correlation of critical heat flux for forced convection boiling in vertical uniformly heated round tubes", Heat Mass Tran, 21, pp.1527~1542 (1978)
  13. S. J. Kline, "Describing uncertainties in single-sample experiments", Mech. Eng, 75, pp.3~8 (1953)
  14. I.L. Pioro, "Comparison of CHF measurements in R-134a cooled tubes and the water CHF look-up table", International Journal of Heat and Mass Transfer, 44(1), pp.73~88 (2001), Elsevier BV[DOI]
  15. Burnout Conditions for Single Rod in Annular Geometry, Water at 600 to 1400 Psia, "E. Janssen", ~, pp.GEAP-3899 (1963)
  16. I.L Pioro, "Experimental evaluation of the limiting critical quality values in circular and non-circular flow geometries", Nuclear Engineering and Design, 190(3), pp.317~339 (1999), Elsevier BV[DOI]
  17. D.C. Groeneveld, "The 2006 CHF look-up table", Nuclear Engineering and Design, 237(15-17), pp.1909~1922 (2007), Elsevier BV[DOI]

관련 논문AI추천

An experimental study of the CHF in a square channel with non-uniformly heated rod under low pressure and low flow conditionsChen Lixin, Chen Sen, Li Da, Li Huaqi, Luo XiaoFei, Shi Leitai, Tian Xiaoyan, Zhu Lei한국원자력학회 · Nuclear Engineering and Technology2025 · 제57권 제4호A Method for Critical Heat Flux Prediction in Vertical Round Tubes with Axially Non-uniform Heat Flux Profile심재우한국해양공학회 · 한국해양공학회지2008 · 제22권 제1호Experimental study on the Influence of Heating Surface Inclination Angle on Heat Transfer and CHF performance for Pool BoilingChenglong Wang, Dalin Zhang, G.H. Su, Jian Deng, Panxiao Li, Suizheng Qiu, Wenxi Tian한국원자력학회 · Nuclear Engineering and Technology2022 · 제54권 제1호Numerical investigation of the critical heat fl ux in a 5 × 5 rod bundle with multi-gridLixin Yang, Qian Peng, Shihao Yang, Wei Liu, Xi Chen, Yu Liu, Zemin Shang, Zihao Tian한국원자력학회 · Nuclear Engineering and Technology2022 · 제54권 제5호Water film covering characteristic on horizontal fuel rod under impinging cooling conditionChen Ronghua, Qiu Suizheng, Su G.H., Tian Wenxi, Wang Bowei, Zhang Penghui한국원자력학회 · Nuclear Engineering and Technology2022 · 제54권 제11호Experimental and numerical investigation on the flow and heat transfer behavior of double-side heating once-through steam generator with helical tubesFu Shengwei, Jiang Jiahang, Jiao Meng, Zhao Xinwen한국원자력학회 · Nuclear Engineering and Technology2025 · 제57권 제6호

이 간행물의 다른 논문 (39건)