Nuclear Engineering and Technology

Korean Nuclear Society (한국원자력학회)
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A REVIEW OF INHERENT SAFETY CHARACTERISTICS OF METAL ALLOY SODIUM-COOLED FAST REACTOR FUEL AGAINST POSTULATED ACCIDENTS

  • Received : 2015.03.09
  • Accepted : 2015.03.10
  • Published : 2015.04.25
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Abstract

The thermal, mechanical, and neutronic performance of the metal alloy fast reactor fuel design complements the safety advantages of the liquid metal cooling and the pool-type primary system. Together, these features provide large safety margins in both normal operating modes and for a wide range of postulated accidents. In particular, they maximize the measures of safety associated with inherent reactor response to unprotected, doublefault accidents, and to minimize risk to the public and plant investment. High thermal conductivity and high gap conductance play the most significant role in safety advantages of the metallic fuel, resulting in a flatter radial temperature profile within the pin and much lower normal operation and transient temperatures in comparison to oxide fuel. Despite the big difference in melting point, both oxide and metal fuels have a relatively similar margin to melting during postulated accidents. When the metal fuel cladding fails, it typically occurs below the coolant boiling point and the damaged fuel pins remain coolable. Metal fuel is compatible with sodium coolant, eliminating the potential of energetic fuel-coolant reactions and flow blockages. All these, and the low retained heat leading to a longer grace period for operator action, are significant contributing factors to the inherently benign response of metallic fuel to postulated accidents. This paper summarizes the past analytical and experimental results obtained in past sodium-cooled fast reactor safety programs in the United States, and presents an overview of fuel safety performance as observed in laboratory and in-pile tests.

Keywords

References

  1. The Generation IV International Forum (GIF) [Internet]. OECD Nuclear Energy Agency, Available from: https://www.gen-4.org/gif/jcms/c_9260/public.
  2. The International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO) [Internet]. International Atomic Energy Agency. Available from: http://www.iaea.org/INPRO/.
  3. L.C. Walters, B.R. Seidel, J.H. Kittel, Performance of metallic fuels and blankets in liquidemetal fast breeder reactors, Nucl. Technol. 65 (1984) 179-231. https://doi.org/10.13182/NT84-A33408
  4. D. Crawford, D. Porter, S. Hayes, Fuels for sodium-cooled fast reactors: US perspective, J. Nucl. Mater. 371 (2007) 202-231. https://doi.org/10.1016/j.jnucmat.2007.05.010
  5. Y. Chang, Technical rationale for metal fuel in fast reactors, Nucl. Eng. Technol. 39 (2007) 161-170. https://doi.org/10.5516/NET.2007.39.3.161
  6. H.P. Planchon, et al., Implications of the EBR-II inherent safety demonstration test, Nucl. Eng. Des. 101 (1987) 75-90. https://doi.org/10.1016/0029-5493(87)90152-X
  7. C.M. Walter, L.R. Kelman, The interaction of iron with molten uranium, J. Nucl. Mater. 20 (1966) 314. https://doi.org/10.1016/0022-3115(66)90044-4
  8. P.R. Betten, J.H. Bottcher, B.R. Seidel, Eutectic penetration times in irradiated EBR-II driver fuel elements, Trans. Am. Nucl. Soc. 45 (1983) 300.
  9. B.R. Seidel, Metallic fuel cladding eutectic formation during postirradiation heating, Trans. Am. Nucl. Soc. 34 (1980) 210.
  10. C.E. Lahm, J.F. Koenig, P.R. Betten, J.H. Bottcher, W.K. Lehto, B.R. Seidel, EBR-II driver fuel qualification for loss of flow and loss of heat sink tests without scram, Nucl. Eng. Des. 101 (1987) 25. https://doi.org/10.1016/0029-5493(87)90147-6
  11. C.E. Dickerman, et al., TREAT sodium loop experiments on performance of unbonded, unirradiated EBR-II Mark I fuel elements, Nucl. Eng. Des. 12 (1970) 381-390. https://doi.org/10.1016/0029-5493(70)90052-X
  12. W.R. Robinson, et al., Integral fast reactor safety tests M2 and M3 in TREAT, Trans. Am. Nucl. Soc. 50 (1985) 352.
  13. A.E. Wright, et al., Recent Metal Fuel Safety Tests in TREAT, Proceedings of the ANS/ENS International Conference of the Science and Technol. of Fast Reactor Safety, CONF-86050, Guernsey, England, May 1986.
  14. T.H. Bauer, et al., Behavior of uraniumefissium fuel in TREAT transient overpower tests, Trans. Am. Nucl. Soc. 53 (1986) 306.
  15. W.R. Robinson, et al., First TREAT transient overpower tests on UePueZr Fuel: M5 and M6, Trans. Am. Nucl. Soc. 55 (1987) 418.
  16. T.H. Bauer, G.R. Fenske, J.M. Kramer, Cladding Failure Margins for Metallic Fuel in the Integral Fast Reactor, Transactions of SMiRT-9, Lausanne, Switzerland, August 1987.
  17. B.W. Spencer, J.F. Marchaterre, Scoping Studies of Vapor Behavior during a Severe Accident in a Metal-fueled Reactor, Proc. Intl. Topical Mtg. on Fast Reactor Safety, Knoxville, TN, CONF-850410 vol. 1, April 1985, p. 151.
  18. T. Ogata, Y.S. Kim, A.M. Yacout, Metal fuel modeling and simulation, in: Comprehensive Nuclear Materials, Elsevier, 2011 (Chapter 75).

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