Nuclear Engineering and Technology

Korean Nuclear Society (한국원자력학회)
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Transient Critical Heat Flux Under Flow Coastdown in a Vertical Annulus With Non-Uniform Heat Flux Distribution

  • Published : 2002.08.01
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Abstract

An experimental study on transient critical heat flux (CHF) under flow coastdown has been performed for the water flow in a non-uniformly heated vertical annulus under low flow and a wide range of pressure conditions. The objectives of this study are to systematically investigate the effect of the flow transient on the CHF and to compare the transient CHF with steady-state CHF The transient CHF experiments have been performed for three kinds of flow transient modes based on the coastdown data of a nuclear power plant reactor coolant pump. At the same inlet subcooling, system pressure and heat flux, the effect of the initial mass flux on the critical mass flux can be negligible. However, the effect of the initial mass flux on the time-to- CHF becomes large as the heat flux decreases. The critical mass flux has the largest value for slow flow reduction rate. There is a pressure effect on the ratio of the transient CHF data to steady-state CHF data. Except under low system pressure conditions, the flow transient CHF was revealed to be conservative compared with the steady-state CHF data. Bowling CHF correlation and thermal hydraulic system code MARS show promising results for the prediction of CHF occurrence .

Keywords

References

  1. S. H. Chang and W. P. Baek, 'Perspective on Critical Heat Flux Research for Nuclear Reactors,' Proc. of NTAHAS 98: First Korea-Japan Symposium on Nuclear Thermal Hydraulic and Safety, Pusan, Korea, October 21-24 (1998)
  2. T. Iwamura, 'Transient Burnout Under Rapid Row Reduction Condition,' J. of Nucl. Sci. and Tech., 24(10), 811-820 (1987) https://doi.org/10.3327/jnst.24.811
  3. G. P. Celata, M. Cumo et al., 'CHF Behavior During Pressure, Power and/or Flow Rate Simultaneous Variations,' Int. J. Heat Mass Transfer, 34[3], 723-738 (1991) https://doi.org/10.1016/0017-9310(91)90120-4
  4. M. Cumo, F. Fabrizi and G. Palazzi, 'Transient Critical Heat Flux in Loss-of-Flow-Accidents (L.O.F.A),' Int. J. Multiphase Row, 4, 497-509 (1978) https://doi.org/10.1016/0301-9322(78)90010-1
  5. S. H. Chang, K. W. Lee and D. C. Groeneveld, 'Transient-Effects Modeling of Critical Heat Flux,' Nucl. Eng. Design, 113, 51-57 (1989) https://doi.org/10.1016/0029-5493(89)90295-1
  6. J. C. M. Leung, 'Critical Heat Flux Under Transient Conditions: A Literature Survey,' ANL Report, ANL-78-39, NUREG/CR-0056 (1978)
  7. S. H. Chang and W. P. Baek, 'Critical Heat Flux - Fundamentals and Applications,' Chungmoongak Pub. Co., Seoul (in Korean) (1997)
  8. S. K. Moon, S. Y., Chun, H. J., Chung et al., 'Effect of Axial Heat Flux Distributions on Critical Heat Flux under Low Flow and a Wide Range of System Pressures with Vertical Annulus,' Proc. of 8th Int. Conf. on Nuclear Engineering, April 2-6, 2000, Baltimore, MD USA (2000)
  9. ANSI/ASME PTC 19.1, 'ASME Performance Test Codes, Supplement on Instruments and Apparatus, Part 1, Measurement Uncertainty' (1985)
  10. K. Mishima and M. Ishii, 'Flow Regime Transition Criteria for Upward Two-Phase Flow in Vertical Tubes,' Int. J. Heat Mass Transfer, 27[5]. 723-737 (1984) https://doi.org/10.1016/0017-9310(84)90142-X
  11. R. W. Bowring, 'A New Mixed Row Cluster Dryout Correlation for Pressures in the Range 0.6-15.5 MN/$M^2$ (90-2500 psia) - for Use in a Transient Blowdown Code,' Paper C217/77, Presented at Conf. On Heat and Fluid Row in Water Reactor Safety, IMechE, Manchester, Sep. 13-15 (1977)
  12. S. Y. Chun, H. J. Chung, S. K. Moon et al., 'Effect of Pressure on Critical Heat Flux in Uniformly Heated Vertical Annulus under Low Flow Conditions,' Nucl. Eng. Design, 203, 159-174 (2001) https://doi.org/10.1016/S0029-5493(00)00307-1
  13. W. J. Lee, B. D. Chung, J. J. Jeong and K. S. Ha, 'Development of a Multi-Dimensional Realistic Thermal-Hydraulic System Analysis Code, MARS 1.3 and Its Verification,' KAERI/TR-1108/98 (1998)
  14. J. H. Chun, H. J. Chung, S. Y. Chun and U. C. Lee, 'Assessment of MARS 1. 4 Dryout Model Using KAERI Annular CHF Test,' Proc. of the Korean Society of Mechanical Engineering Autumn Meeting (1999)
  15. J. H. Chun , W. J. Lee, and U. C. Lee, 'Mechanistic Modelling of Annular Film Dryout in Annulus Geometry in MARS Code,' Proc. of the Second Japan-Korea Symposium in Nuclear Thermal Hydraulics and Safety, Fukuoka, Japan (2000)
  16. K. Y. Choi, S. K. Moon, S. Y. Chun and J. D. Jackson, 'Prediction of Critical Heat Flux in a Non-Uniformly Heated Vertical Annulus Under Flow Transients Using an Area Split Approach,' Proc. of '2001: A Nuclear Odyssey, ANS/HPS Conf., Texas, USA (2001)
  17. P. L. Kirillov and I. P. Smogalev, 'Calculation of Heat Transfer Crisis for Annular Two-Phase Flow of a Steam-Liquid Mixture Through an Annular Channel,' AECL-4752 (1974)