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Journal of Nuclear Science and Technology Volume 25, 1988 - Issue 2
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Short Note

Critical Heat Flux for Rod Bundle under High-Pressure, Low-Flow and Mixed Inlet Conditions

Hiroshige KUMAMARU Japan Atomic Energy Research Institute, Tokai-mura, Ibaraki-ken, 319-11
,
Yasuo KOIZUMI Japan Atomic Energy Research Institute, Tokai-mura, Ibaraki-ken, 319-11
&
Kanji TASAKA Japan Atomic Energy Research Institute, Tokai-mura, Ibaraki-ken, 319-11
Pages 207-209 | Received 04 Nov 1987, Published online: 15 Mar 2012

REFERENCES

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  • Kumamaru , H. 1987 . Investigation of pre- and post-dryout heat transfer of steam-water two-phase flow in a rod bundle . Nucl. Eng. Des. , 102 : 71 – 84 .
  • Koizumi , Y. 1987 . Post-dryout heat transfer of high-pressure steam-water two-phase flow in single rod channel and multi rod bundle . ibid. , 99 : 157 – 165 .
  • Tasaka , K. “ ROSA-III experimental program for BWR LOCA/ECCS integral system tests, To be published in ” . In JAERI Report.
  • Adachi , H. 1982 . ROSA-III experimental program for PWR LOCA/ECCS integral tests . JAERI , 1277
  • Biasi , L. 1967 . Studies on burnout, Part 3—A new correlation for round ducts and uniform heating and its comparison with world data . Ene. Nucl. , 14 : 530 – 537 .
  • Bowring , R. W. 1977 . “ A new mixed flow cluster dryout correlation for pressures in the range 0.6—15.5 MN/m2 (90-2250 psia) for use in a transient blowdown code ” . In “Heat and Fluid Flow in Water Reactor Safety” 175 Inst. of Mech. Engineers. .
  • Katto , Y. 1978 . A generalized correlation of critical heat flux for the forced convection boiling in vertical uniformly heated round tubes . Int. J. Heat Mass Transfer , 21 : 1527 – 1542 .
  • idem. 1981 . On the heat-flux/exit-quality type correlation of CHF of forced convection boiling in uniformly heated vertical tubes . ibid. , 24 : 533 – 539 .
  • Ranson , V. H. 1985 . RELAP5/MOD2 code manual, volume 1: Code structure, systems models and solution methods . NUREG/CR-4312, EGG-2396 ,
  • Liles , D. 1984 . TRAC-PF1/MOD1; An advanced best-estimate computer program for pressurized water reactor thermal-hydraulic analysis LANL, Draft

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