Critical power analysis with mechanistic models for nuclear fuel bundles, (I): Models and verifications for boiling water reactor application

Journal of Nuclear Science and Technology

Volumn 39, Issue 1, 2002, Pages 40-52

  Naitoh, Masanori ^a   Ikeda, Takashi ^a   Nishida, Koji ^b   Okawa, Tomio ^c   Kataoka, Isao ^c  

^a Nuclear Power Engineering Corp   (Japan)

^b HITACHI LTD   (Japan)

^c OSAKA UNIVERSITY   (Japan)

Author keywords

BWR type reactors; Critical power; Dryout; Fuel bundle; Fuel element clusters; Mechanistic model; Parallel computation; Spacer; Subchannel analysis

Indexed keywords

BOUNDARY CONDITIONS; NUCLEAR FUEL ELEMENTS; PARALLEL PROCESSING SYSTEMS; STATISTICAL METHODS;

CRITICAL POWER ANALYSIS; NUCLEAR FUEL BUNDLES;

BOILING WATER REACTORS;

EID: 0036196935     PISSN: 00223131     EISSN: None     Source Type: Journal    
DOI: 10.1080/18811248.2002.9715155     Document Type: Article

References (25)

1. A. Tomiyama, O. Yokomizo, “Spacer effects on film flow in BWR fuel bundle,” J. Nucl. Sci. Technol, 25[2], 94 (1988).
2. K. Nishida, T. Kanazawa, O. Yokomizo, Y. Masuhara, “Spacer effect on liquid film flow and critical power in BWR fuel bundles,” J. Nucl. Sci. Technol., 31[3], 213 (1994).
3. S. Morooka, K. Shirakawa, Y. Yamamoto, et al, “Development of high performance BWR fuel bundle—Thermal hydraulic performance of CYCLONE spacer,” Nihon-Genshiryoku-Gakkai Shi (J. At. Energy Soc. Jpn), 38[9], 771 (1996), [in Japanese].
4. A. Tomiyama, O. Yokomizo, Y. Yoshimoto, S. Sugawara, “Method of critical power prediction based on film flow model coupled with subchannel analysis,” J. Nucl. Sci. Technol., 25[12], 914 (1988).
5. T. Sakai, H. E. C. Rummens, S. Sugawara, “FIDAS: A three-fluid subchannel code for dryout prediction in rod bundles,” Proc. 10 th Annu. Conf. of Canadian Nuclear Society, Ottawa, Canada, June 5-8, 1989, (1989).
6. S. Sugawara, T. Sakai, K. Watanabe, H. E. C. Rummens, “Subchannel analysis by the FIDAS code based on the three-fluid model,” Proc. 4th Int. TopicalMtg. on Nucl. Reactor Thermal-Hydraulics (NURETH-4), Karlsruhe, Germany, Oct. 10-13, 1989, Vol. 1, p. 560 (1989).
7. S. Morooka, Y. Yamamoto, “A prediction method for critical power of new BWR fuel assemblies (2nd report),” Preprints 1997 Fall Meeting of At. Energy Soc. Jpn., Okinawa, Japan, Oct. 14-17, 1997, G27, (1997), [in Japanese].
8. T. Kanazawa, K. Nishida, T. Nagayoshi, et al, “Critical power evaluation with consideration of differences in droplet behavior due to spacer shape influence,” J. Nucl. Sci. Technol., 32[4], 301 (1995).
9. G. B. Wallis, “Annular two phase flow, part I: A simple theory,” J. Basic Eng., 92, 59(1970).
10. R. T. Lahey Jr., F. J. Moody, The Thermal-hydraulics of a Boiling Water Nuclear Reactor, ANS Monograph, 122 (1977).
11. N. Zuber, J. A. Findley, “Average volumetric concentration in two phase flow system,” J. Heat Transfer, 87,453 (1965).
12. G. E. Dix, Vapor Void Fractions for Forced Convection with Subcooled Boiling at Low Flow Rates, GE Report NEDO-10491,(1971).
13. R. T. Lahey Jr., B. S. Shiralkar, D. W. Radcliffe, “Mass flux and enthalpy distribution in a rod bundle for single- and two-phase flow conditions,” J. Heat Transfer, 93, 197 (1971).
14. A. Inoue, M. Futakuchi, M. Yagi, et al, “Void fraction distribution in a Boiling Water Reactor fuel assembly and the evaluation of subchannel analysis codes,” Nucl. Technol., 112[3], 388(1995).
15. P. B. Whalley, P. Hutchinson, P. W. James, “The calculation of critical heat flux in complex situations using an annular flow model,” 6th Int. Heat Transfer Conf., Toronto, Canada, Vol. 5, p. 65 (1978).
16. Y. Katto, “Prediction of critical heat flux for annular flow in tubes taking into account of critical liquid film thickness concept,” Int. J. Heat Mass Transfer, 27[6], 883 (1984).
17. M. Sadatomi, Y. Sato, “Two-phase flow in multiple channels (2nd report, Flow characteristics),” J. JSME, Ser. B, 53[487], 731 (1987), [in Japanese].
18. J. Wurtz, An Experimental and Theoretical Investigation of Annular Steam-water Flow in Tubes and Annuli at 30 to 90 bar, RISO Report No. 372, (1978).
19. T. Mitsutake, M. Kitamura, H. Hayashi, et al, “Proving test for critical power performance of BWR high burnup 8x8 fuel assembly and evaluation of subchannel code,” Proc. 5th Int. Conf. Nucl. Eng. (ICONE5), Nice, France, May 25-29, 1997, Paper No. ICONE5-2315, (1997).
20. M. Naitoh, Y. Ando, H. Ujita, et al., “Development of the accident simulation system ‘IMPACT’ for light water nuclear power plants,” Nihon-Genshiryoku-Gakkai Shi (J. At. Energy Soc. Jpn), 41[3], 174 (1999), [in Japanese].
21. T. Tanaka, K. Amano, M. Yamakawa, et al., “Numerical model of pulverized coal particle dispersion,” Proc. 31st National Heat Transfer Symp. of Japan, Sapporo, Japan, May 18-20, 1994, Vol. 2, F222, 688 (1994), [in Japanese].
22. T. Ueda, “Droplet entrainment rate and droplet diameter in annular two-phase flow,” Bull. JSME, 45[389], 127 (1979), [in Japanese].
23. T. Nagayoshi, K. Nishida, T. Kawasaki, Y. Bessho, “Development of subchannel analysis methods in BWRs: Fundamental experiment on turbulence downstream of fuel spacer model,” Preprints 1994 Fall Meeting of At. Energy Soc. Jpn., Sapporo, Japan, Sep. 28-30, 1994, A35, (1994), [in Japanese].
24. T. Mitsutake, H. Terasaka, K. Yoshimura, et al., “Subchannel analysis of a critical power test, using simulated BWR 8x8 fuel assembly,” Nucl. Eng. Des., 122,235 (1990).
25. C. F. Fighetti, D. G. Reddy, Parametric Study of CHF Data, EPRI-NP-2609, Vol. 3, part. 2, (1982).