Uranium-zirconium hydride fuel

D. Olander, K. Konashi, M. Yamawaki

Research output: Chapter in Book/Report/Conference proceedingChapter

4 Citations (Scopus)

Abstract

The properties of the two-phase uranium-zirconium hydride U0.3ZrH1.6 pertaining to its performance as a nuclear fuel for uranium-zirconium light-water reactors (LWRs) are reviewed. Much of the available data come from the Space Nuclear Auxiliary Power (SNAP) program of four decades ago and from the more restricted database prepared for the Training, Research, Isotopes, General Atomics (TRIGA) research reactors some three decades back. The fuel's transport, mechanical, thermal, and chemical properties are summarized. A principal difference between oxide and hydride fuels is the high thermal conductivity of the latter. This feature greatly decreases the temperature drop over the fuel during operation, thereby reducing the release of fission gases to the fraction due only to recoil. However, very unusual early swelling due to void formation around the uranium particles has been observed in hydride fuels. Avoidance of this source of swelling limits the maximum fuel temperature to ∼650°C. To satisfy this temperature limitation, the fuel-cladding gap needs to be bonded with a liquid metal (LM) instead of helium. Because the former has a thermal conductivity ∼100 times larger than the latter, there is no restriction on gap thickness as there is in helium-bonded fuel rods. This opens up the possibility of initial gap sizes large enough to significantly delay the onset of, or even avoid, pellet-cladding mechanical interaction (PCMI). The LM bond permits operation of the fuel at current LWR linear-heat-generation rates without exceeding any design constraint. The behavior of hydrogen in the fuel during operation is the source of phenomena that are absent in oxide fuels. Because of the large heat of transport (thermal diffusivity) of H in ZrHx, redistribution of hydrogen in the temperature gradient in the fuel pellet changes the initial H/Zr ratio of 1.6 to ∼1.45 at the center and ∼1.70 at the periphery. Because the density of the hydride decreases with increasing H/Zr ratio, H redistribution subjects the interior of the pellet to a tensile stress while the outside of the pellet is placed in compression. The resulting stress at the pellet periphery is sufficient to overcome the tensile stress due to thermal expansion in the temperature gradient and to prevent radial cracking that is a characteristic of oxide fuel. Several mechanisms for reduction of the H/Zr ratio during irradiation are identified. The first is the transfer of impurity oxygen in the fuel from Zr to rare-earth oxide fission products (FPs). The second is the formation of metal hydrides by these same FPs. The third is by loss to the plenum as H2.The review of the fabrication method for the hydride fuel suggests that the cost of its production, even on a large scale, may be significantly higher than that of oxide fuel fabrication.

Original languageEnglish
Title of host publicationComprehensive Nuclear Materials
PublisherElsevier Ltd
Pages313-357
Number of pages45
Volume3
ISBN (Print)9780080560335
DOIs
Publication statusPublished - 2012 Dec 1

Keywords

  • Advanced fuel
  • Hydride fuel
  • Irradiation testing
  • Liquid-metal bond

ASJC Scopus subject areas

  • Energy(all)

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  • Cite this

    Olander, D., Konashi, K., & Yamawaki, M. (2012). Uranium-zirconium hydride fuel. In Comprehensive Nuclear Materials (Vol. 3, pp. 313-357). Elsevier Ltd. https://doi.org/10.1016/B978-0-08-056033-5.00061-6