Advertisement

Fuel for the SGHWR

  • D. O. Pickman
  • J. H. Gittus
  • K. M. Rose

Abstract

The Steam Generating Heavy Water Reactor system (SGHWR) was selected for development in the United Kingdom as an alternative to the Commercial Advanced Gas-Cooled Reactor (CAGR) in the late 1950’s. As an alternative, it was required to have diverse technical features such as coolant, moderator and pressure containment.

Keywords

Fuel Element Pressure Tube Fuel Cladding Power Cycling Steam Drum 
These keywords were added by machine and not by the authors. This process is experimental and the keywords may be updated as the learning algorithm improves.

Preview

Unable to display preview. Download preview PDF.

Unable to display preview. Download preview PDF.

References

  1. 1.
    Nuclear Reactor Systems for Electricity Generation, HMSO, London, July, 1974.Google Scholar
  2. 2.
    Bradley, N., Dawson, D. J. and Johnson, F. G., Engineering design of SGHWR’s, Proceedings BNES Conference, “Steam Generating and Other Heavy Water Reactors,” London, 1968, Page 11.Google Scholar
  3. 3.
    Smith, D. R. and Phillips, J. L., “The SGHWR — Design and Operational Experience,” Nuclex 1975, Technical Meeting 3/13.Google Scholar
  4. 4.
    Pickman, D. O., “Design of Fuel Elements,” Nuclear Engineering Design 21, 1972, Page 303.CrossRefGoogle Scholar
  5. 5.
    Pickman, D. O., Willey, D. H. and Eldred, V. W., Proceedings BNES Conference, “Nuclear Fuel Performance,” Paper 51, London, 1973.Google Scholar
  6. 6.
    Collins, D. A., Hargreaves, R. and Hughes, H., Proc. BNES Conference, “Nuclear Fuel Performance,” Paper 49, London, 1973.Google Scholar
  7. 7.
    Hargreaves, R., BNES Journal, to be published, 1976.Google Scholar
  8. 8.
    Trowse, F. W., Garlick, A. and Sumerling, R., “Nodular Corrosion of Zircaloy–2 and Some Other Zirconium Alloys in SGHWR and Related Environments,” Paper to be presented at ASTM Symposium, Zirconium in the Nuclear Industry, Quebec City, August, 1976.Google Scholar
  9. 9.
    Tyzack, C. and Sheppard, M. F., Paper to be presented at ASTM Symposium, Zirconium in the Nuclear Industry, Quebec City, August, 1976.Google Scholar
  10. 10.
    Pickman, D. O., “Fuel Performance in the Prototype SGHWR Power Station,” UKAEA TRG Report 1943 (S), HMSO, London December, 1969.Google Scholar
  11. 11.
    Bond, G. G., Cordall, D., Cornell, R. M., Fox, W. N., Garlick, A. and Howl, D. A., “SGHWR Fuel Performance Under Power Ramp Conditions,” BNES, 1976.Google Scholar
  12. 12.
    Garlick, A., J. Nuclear Materials 49, Page 209, 1973.ADSCrossRefGoogle Scholar
  13. 13.
    Gittus, J. H., Howl, D. A. and Hughes, H., “Theoretical Modeling of Nuclear Fuel Performance,” Report of Work at the UKAEA Springfields Laboratories, TRG Report 2743 (S), 1975.Google Scholar
  14. 14.
    Gittus, J. H., Howl, D. A. and Hughes, H., Nuclear Applied Tech. 9, PP 40–46, 1970.Google Scholar
  15. 15.
    Gittus, J. H., Nuclear Engineering Des. 18, PP 69–82, 1972.CrossRefGoogle Scholar
  16. 16.
    Gittus, J. H., Creep Viscoelasticity and Creep-Fracture in Solids, London: Elseviers Applied Science Publishers Ltd; New York: Halstead, Division of Wiley, 1975.Google Scholar
  17. 17.
    Gittus, J. H., ORNL TM 2857. See also, Gittus, J. H., Nuclear Engineering Design 28, PP 252–256, 1974.Google Scholar
  18. Gittus, J. H., ORNL TM 2857. See also, Gittus, J. H., Nuclear Engineering Design 28, PP 252–256, 1974.CrossRefGoogle Scholar
  19. 18.
    Gittus, J. H., Proceedings International Conference Metallurgy of Reactor Fuel Elements, CEGB, Berkeley Nuclear Laboratories, Gloucestershire, England, PP 369–373, 1973.Google Scholar
  20. 19.
    Smith, D. R., Bradley, N., Hicks, D. and Cowan, A., “SGHWR Loss-of-coolant Accidents,” Nuclex 1975, Technical Meeting 5/09.Google Scholar
  21. 20.
    Scatena, G. J., “Fuel Cladding Embrittlement During a Loss-of-coolant Accident,” NEDO 10674, October, 1972.Google Scholar
  22. 21.
    USNRC, Technical Information Division, “The Role of Fission Gas Release in Reactor Licensing,” NUREG-75/077.Google Scholar
  23. 22.
    Hindle, E. D., UKAEA RFL Springfields Internal DocumentGoogle Scholar
  24. 23.
    Clay, B. D., Redding, G. B., “Creep Properties of Alpha phase Zircaloy-2 Cladding Relevant to the Loss-of-Coolant Accident,” CEGB Report RD/B/N3187, March, 1975.Google Scholar
  25. 24.
    Hardy, D. G., “High Temperature Expansion and Rupture Behavior of Zircaloy Tubing,” ANS Topical Meeting on Water Reactor Safety, Page 254, 1973.Google Scholar
  26. 25.
    Clendenning, W. R., “Primary and Secondary Creep Properties for Zircaloy Cladding at Elevated Temperatures of Interest in Accident Analyses,” Third International Conference Str. Mech. Reactor Tech. C2/6, London, September, 1975.Google Scholar
  27. 26.
    Gittus, J. H., “Creep of Two-phase Material,” Presented at the Royal Society’s Discussion Meeting on Creep in Engineering Materials and the Earth, London, The Royal Society, 1977.Google Scholar
  28. 27.
    Hindle, E. D., UKAEA RFL Springfields, internal document.Google Scholar
  29. 28.
    Emmerich, K. M., Juenke, E. F. and White, J. F., “Failure of Pressureized Zircaloy Tubes during Thermal Excursions in Steam in Inert Atmospheres,” ASTM STP 458, PP 252–268, 1969.Google Scholar
  30. 29.
    Hobson, D. O., Osborne, M. E. and Parker, G. W., “Comparison of Rupture Data from Irradiated Fuel Rods and Unirradiated Cladding,” Nuclear Technology, Vol. 11, Page 479, August, 1971.Google Scholar
  31. 30.
    Gittus, J. H., Acta Met. 22, Page 1179 (Equation 12), 1974.CrossRefGoogle Scholar

Copyright information

© Plenum Press, New York 1979

Authors and Affiliations

  • D. O. Pickman
    • 1
  • J. H. Gittus
    • 1
  • K. M. Rose
    • 1
  1. 1.Reactor Fuel Element LaboratoriesUKAEASpringfields, Preston, LancashireUK

Personalised recommendations