Nuclear Engineering and Technology

  • 제37권5호
  • /
  • Pages.479-490
  • /
  • 2005
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  • 1738-5733(pISSN)
  • /
  • 2234-358X(eISSN)
한국원자력학회 (Korean Nuclear Society)
권호 표지

THERMALHYDRAULIC EVALUATIONS FOR A CANFLEX BUNDLE WITH NATURAL OR RECYCLED URANIUM FUEL IN THE UNCREPT AND CREPT CHANNELS OF A CANDU-6 REACTOR

  • 발행 : 2005.10.01
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초록

The thermalhydraulic performance of a CANDU-6 reactor loaded with various CANFLEX fuel bundles is evaluated by the NUCIRC code, which is incorporated with recent models of pressure drop and critical heat flux (CHF) predictions based on high-pressure steam-water tests for the CANFLEX bundle as well as a 37-element bundle. The distributions of channel flow rate, channel exit quality, critical channel power (CCP), and critical power ratio (CPR) for the CANFLEX bundles (with natural or recycled uranium fuel) in the CANDU-6 reactor fuel channel are calculated by the code. The effects of axial and radial heat flux on CCP are evaluated by assuming that the recycled uranium fuel (CANFLEX-RU) has the same geometric data as the natural uranium fuel bundle (CANFLEX-NU), but a different power distribution due to different fuel composition and refueling scheme. In addition, the effects of pressure tube creep and bearing-pad height are examined by comparing various results of uncrept, and $3.3\%\;and\;5.1\%$ crept channels loaded with CANFLEX bundles with 1.4 mm or 1.7 mm high bearing-pads with those of the 37-element bundle. The distributions of the channel flow rate and CCP for the CANFLEX-NU or -RU bundle show a typical trend for a CANDU-6 reactor channel, and the CPRs are maintained above at least 1.444 (NU) or 1.455 (RU) in the uncrept channel. The enhanced CHF of the CANFLEX bundle (particularly with 1.7mm height bearing-pads) produces a higher thermal margin and considerably less sensitivity to CCP reduction due to the pressure tube creep than the 37-element bundle. The CCP enhancement due to the raised bearing-pads is estimated to be about $3\%\~5\%$ for the CANFLEX-NU and $2\%\~6\%$ for the CANFLEX-RU bundle, respectively.

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참고문헌

  1. W.W.R. Inch, H.C. Suk, 'Demonstration Irradiation of CANFLEX in Pt. Lepreau', IAEA Technical Committee Meeting on Fuel Cycle Options for LWRs and HWRs, Victoria, Canada, April 1998
  2. J.Y. Jung, J.S. Jun, M.S. Cho, H.C. Suk, S.D. Lee and H.B. Seo, 'Evaluation of CANFLEX-NU Fuel Performance Irradiated in Wolsong Generation Station #1 and In-Bay Inspection', Proceedings of the KNS Spring Meeting, Korea, May 2004
  3. G.R. Dimmick, W.W.R. Inch, J.S. Jun, H.C. Suk, G.I. Hadaller, R.A. Fortman and R.C. Hayes, 'Full Scale Water CHF Testing of the CANFLEX Bundle', the 6th International Conference on CANDU Fuel, Canadian Nuclear Society, September 1999
  4. L.K.H. Leung, J.S. Jun, D.E. Bullock, W.W.R. Inch and H.C. Suk, 'Dryout Power in a CANFLEX Bundle String with Raised Bearing Pads', the 7th International Conference on CANDU Fuel, Canadian Nuclear Society, September 2001
  5. H.C. Suk, 'Current Status and Future Prospect of CANDU Fuel Research and Development in Korea', 7th International CANDU Fuel Conference Proceedings, September 2001
  6. L.K.H. Leung, D.C. Groeneveld, G.R. Dimmick, D.E. Bullock, and W.W. Inch, 'Critical Heat Flux and Pressure Drop for a CANFLEX Bundle String Inside an Axially Non-Uniform Flow Channel', Proceedings of the 6th International Conference on CANDU Fuel, Canadian Nuclear Society, September 1999
  7. S.S. Doerffer, 'Release of NUCIRC-MOD2.001 for HP/UNIX Use', AECL Memo 00-33000-225-001(internal report), September 2001
  8. R.A. Fortman, G.I. Hadaller, R.C. Hayes and F. Stern, 'Heat Transfer Studies with CANDU Fuel Simulators', the 5th International Conference on Nuclear Engineering, ICONE5, France, May 1997
  9. L. Friedel, 'Improved Friction Pressure Drop Correlations for Horizontal and Vertical Two-Phase Pipe', European Two-Phase Flow Group Meeting, Ispra, Italy, 1979
  10. P. Saha and N. Zuber, 'Point of Net Vapour Generation and Vapour Fraction in Subcooled Boiling', Proceedings of the 5th International Heat Transfer Conference, 1974
  11. M.R. Soulard, 'Evaluation of Two-Phase Multiplier Correlations for Use in NUCIRC Code',TTR-326(internal report), September 1992
  12. J.S. Jun, J.H. Park, and H.C. Suk, 'The CCP Sensitivity Study of Axial Flux Distribution for CANFLEX-NU Fuel Bundle', Proceedings of the KNS Autumn Meeting, Korea, October 1997
  13. J.S. Jun, J.H. Park, and H.C. Suk, 'The CCP Sensitivity Study of CANDU-6 Reactor Channel Loaded with CANFLEX Fuel Bundle', KAERI/TR-893/97(internal report), August 1997
  14. J.S. Jun and H.C. Suk, 'The Thermalhydraulic Characteristics of CANDU-6 Reactor Channel with CANFLEX-RU Fuel Bundle', Proceedings of the KNS Spring Meeting, Korea, May 2003