Nuclear Engineering and Technology

Korean Nuclear Society (한국원자력학회)
권호 표지
DOI QR코드

DOI QR Code

HIGH BURNUP FUEL TECHNOLOGY IN KOREA

  • Published : 2008.02.29
Download PDF
⟨ Previous Next ⟩

Abstract

High bum-up fuel technology has been developed through a national R&D program, which covers key technology areas such as claddings, $UO_2$ pellets, spacer grids, performance code, and fuel assembly tests. New cladding alloys were developed through alloy designs, tube fabrication, out-of-pile test and in-reactor test. The new Zr-Nb tubes are found to be much better in their corrosion resistance and creep strength than the Zircaloy-4 tube, owing to an optimized composition and heat treatment of the new Zr-Nb alloys. A new fabrication technology for large grain $UO_2$ pellets was developed using various uranium oxide seeds and a micro-doping of Al. The uranium oxide seeds, which were added to $UO_2$ powder, were prepared by oxidizing and heat-treating scrap $UO_2$ pellets. A $UO_2$ pellet containing tungsten channels was fabricated for a thermal conductivity enhancement. For the fuel performance analysis, new high burnup models were developed and implemented in a code. This code was verified by an international database and our own database. The developed spacer grid has two features of contoured contact spring and hybrid mixing vanes. Mechanical and hydraulic tests showed that the spacer grid is superior in its rodsupporting, wear resistance and CHF performance. Finally, fuel assembly test technology was also developed. Facilities for mechanical and thermal hydraulic tests were constructed and are now in operation. Several achievements are to be utilized soon by the Korea Nuclear Fuel and thereby contribute to the economy and safety of PWR fuel in Korea

Keywords

References

  1. Y.H. Jeong, H.G. Kim, D.J. Kim, B.K. Choi, J.H. Kim, 'Influence of Nb concentration in the -matrix on the corrosion behavior of Zr-xNb binary alloys', J. Nucl. Mater., 323, 72 (2003) https://doi.org/10.1016/j.jnucmat.2003.08.031
  2. H.G. Kim, Y.H. Jeong, T.H. Kim, 'Effect of phase, precipitate and Nb-concentration in matrix on corrosion and oxide characteristics of Zr-xNb alloys', J. Nucl. Mater., 317, 1 (2003) https://doi.org/10.1016/S0022-3115(02)01676-8
  3. Y.H. Jeong. K.O. Lee, H.G. Kim, 'Correlation between microstructure and corrosion behavior of Zr-Nb binary alloys', J. Nucl. Mater., 302, 9 (2002) https://doi.org/10.1016/S0022-3115(02)00703-1
  4. J.Y. Park, B.K. Choi, Y.H. Jeong, Y.H. Jung, 'Corrosion behavior of Zr alloys with a high Nb content', J. Nucl. Mater., 340, 237 (2005) https://doi.org/10.1016/j.jnucmat.2004.12.003
  5. H.G. Kim, J.Y. Park, Y.H. Jeong, 'Phase boundary of the Zr-rich region in commercial grade Zr-Nb alloys', J. Nucl. Mater., 347, 140 (2005) https://doi.org/10.1016/j.jnucmat.2005.08.008
  6. V.F. Urbanic, B.D. Warr, A. Manolescu, C.K. Chow, M.W. Shanahan, 'Oxidation and deuterium uptake of Zr-2.5Nb pressure tube in CANDU-PHW reactors', Zirconium in the Nuclear Industry, ASTM STP 1023, 20 (1989)
  7. V.F. Urbanic, R.W. Gilbert, 'Effect of microstructure on the corrosion of Zr-2.5%Nb alloys', Proc. of IAEA Technical Committee Meeting on Fundamental Aspects of Corrosion on Zirconium Based Alloys in Water Reactor Environments, International Atomic Energy Agency, Vienna, Italy, IWGFPT/34, p.262 (1990)
  8. J.H. Kim, M.H. Lee, B. K. Choi, Y.H. Jeong, 'Deformation and thermal quench behavior of HANA cladding in LOCA condition', Proc. of 2005 Water Reactor Fuel Performance Meeting, Kyoto, Japan, Oct. 2-6, p.178 (2005)
  9. J.H. Kim, M.H. Lee, B. K. Choi, Y.H. Jeong, 'Deformation of Zircaloy-4 cladding during a LOCA transient up to 1200oC', Nucl. Eng. and Des., 234, 157 (2004) https://doi.org/10.1016/j.nucengdes.2004.08.019
  10. H. Assmann and H. Bairiot, process and Product Control of Oxide Powder and Pellets for Reactor Fuel Application, in: Guidebook on Quality Control of Water Reactor Fuel, Tech. Report Series No. 221, IAEA, Vienna (1983)
  11. J.A. Turnbull, J. Nucl. Mater. 50 (1974) 62 https://doi.org/10.1016/0022-3115(74)90061-0
  12. Y. Harada, J. Nucl. Mater. 238 (1997) 237
  13. K.W. Song, K.S. Kim, K.W. Kang, et al., J. Korean Nucl. Soc. 31 (1999) 335
  14. J.B. Ainscough, F. Rigby, S.C. Osborn, J. Nucl. Mater. 52 (1974) 191 https://doi.org/10.1016/0022-3115(74)90167-6
  15. L. Bourgeois, Ph. Dehaudt, C. Lemaignam, et al., J. Nucl. Mater. 297 (2001) 313 https://doi.org/10.1016/S0022-3115(01)00626-2
  16. B.E. Ingleby, K. Hand, in: c, Advances in Ceramics vol. 17, I.J. Hastings (eds), Amer. Ceramic Soc., 1986, p. 57
  17. Y. Harada and S. Doi, J. Nucl. Sci. Technol., 35 (1998) 411 https://doi.org/10.3327/jnst.35.411
  18. G.A. Wood and C.P. Perkins, in: Advances in Pellet Technology for Improved Performance at High Burnup, IAEA-TECDOC-1036, IAEA, Vienna, 1998, p. 49
  19. K.W. Song, K.S. Kim, K.W. Kang, Y.H. Jung, J. Nucl. Mater. 317 (2003) 204 https://doi.org/10.1016/S0022-3115(03)00080-1
  20. Y.W. Rhee, K.S. Kim, K.W. Song, Thermochimica Acta 455 (2007) 80 https://doi.org/10.1016/j.tca.2006.11.024
  21. C.B. Lee, Y.S. Yang, Y.M. Kim, D.H. Kim, J.G. Bang, Y.H. Jung, 'High Burnup UO2 Fuel Rod Performance Code INFRA', Proc. of the 2004 International Meeting on LWR Fuel Performance, Orlando, Florida, Sep. 19-22, 2004
  22. C.B. Lee, Y.S. Yang, 'INFRA Code Results of FUMEXII', IAEA CRP 'Improvement of models used for fuel behavior simulation (FUMEX-II), IAEA, Vienna, Dec. 5- 8, 2005
  23. C.B. Lee, D.H. Kim, Y.M. Kim, Y.S. Yang, S.K. Kim, Y.H. Jung, Y.B. Chun, H.S. Seo, E.P. Lee, H.K. Lee, H.M. Kwon. Y.H. Jung, D.K. Min, J.I. Kim, O.H. Kim, H.D. Chae, 'Post-irradiation Examination of High Burnup $UO_2$ Fuel', Proc. of the 2004 International Meeting on LWR Fuel Performance, Orlando, Florida, Sep. 19-22, 2004
  24. D. H. Kim et al., 'Post-Irradiation Annealing Test of High Burnup $UO_2$ Fuel', 2005 Water Reactor Fuel Performance Meeting, Kyoto, Japan, Oct. 2-6, 2005
  25. S. Ravel, et al., 'Partitoning of grain boundary and matrix gas inventories: Results obtained using the ADAGIO facility', Fission Gas Behavior in Water Reactor Fuels, Cadarache, France, 2002
  26. Averty, X. et. al., 'Tensile tests on ring specimens machined in M5 cladding irradiated 6 cycles' IRSN 2003/50, 2003
  27. Daum, R. S. et. al., 'On the Embrittlement of Zircaloy-4 Under RIA-Relevant Conditions' Zirconium in the Nuclear Industry : Thirteenth International Symposium, ASTM STP 1423, PA, pp. 702-719, 2002
  28. K.H. Yoon, K.N. Song, T.H. Chun, D.S. Oh, W.K. In, H.K. Kim, H.S. Kang, and Y.H. Jung, 'Spacer Grid for Nuclear Fuel Assemblies with Grid Springs Maintaining Conformal Contact with Fuel Rods and Enlarged Elastic Range,' US Patent US 6707872 B2, 2004
  29. K.H. Yoon, K.N. Song, T.H. Chun, D.S. Oh, W.K. In, H.K. Kim, H.S. Kang, and Y.H. Jung, 'Spacer Grid with H-spring for Fuel Rods for Use in Nuclear Reactor Fuel Assemblies,' US Patent US 6167105, 2000
  30. H.S. Kang, K.N. Song, K.H. Yoon, T.H. Chun, D.S. Oh, W.K. In, H.K. Kim, and Y.H. Jung, 'Side-slotted Nozzle type Double Sheet Spacer Grid for Nuclear Fuel Assemblies,' US Patent US 6744843 B2, 2004
  31. H.S. Kang, K.N. Song, K.H. Yoon, T.H. Chun, D.S. Oh, W.K. In, H.K. Kim, and Y.H. Jung, 'Grid with Nozzletype Coolant Deflecting Channels for Use in Nuclear Reactor Fuel Assemblies,' US Patent US 6130927, 2000
  32. K.N. Song et al., 'Multydisplinary Design of the Spacer Grid Assembly for PWR,' IWMDE2006, Japan, 2006
  33. Frick et al., 'Overview on the Development and Implement of Methodologies to Compute Vibration and Wear of Steam Generator Tubes,' Symposium on Flow Induced Vibration: Vol. 3, Vibration in Heat Exchanger, pp. 149-161, 1984
  34. S.H. Lee, J.Y. Kim, and K.N. Song, Design Improvement of an OPT-H Type Nuclear Fuel Rod Support Grid by Using an Axiomatic Design and an Optimization, Journal of Mechanical Science and Technology, Vol. 21, pp.1191- 1195, 2007 https://doi.org/10.1007/BF03179035
  35. C.H. Shin, Y.J. Choo, S.K. Moon, S.Y. Chun, T.H. Chun, CHF Performance of a Hybrid Mixing vane Grid for Nuclear Fuel Bundle, Proc. 2007 Int. LWR Fuel Performance Mtg, San Francisco, Sep. 30-Oct. 3, 2007
  36. Vishay Measurements Group Inc., 2001, '$StrainSmart^{\circledR}$ User's Manual,' CT, USA
  37. Structural Dynamics Research Corporation, 2002, 'I-DEAS Master Series 9.0,' OH, USA

Cited by

  1. Thermal Conductivity Measurement of Zr-ZrO<sub>2</sub> Simulated Inert Matrix Nuclear Fuel Pellet vol.03, pp.02, 2013, https://doi.org/10.4236/wjnst.2013.32008
  2. Effect of burn-up on the radioactivation behavior of cladding hull materials studied using the ORIGEN-S code vol.298, pp.3, 2013, https://doi.org/10.1007/s10967-013-2543-6