Experimental studies of U-Pu-Zr fast reactor fuel pins in the experimental breeder reactor-ll

Abstract

Argonne National Laboratory’s Integral Fast Reactor (IFR) concept has been under demonstration in the Experimental Breeder Reactor II (EBR-II) since February 1985. Irradiation tests of U-Zr and U-Pu-Zr fuel pins to >15 at. pct burnup have demonstrated their viability as driver fuel prototypes in innovative design liquid metal reactors. A number of technically challenging irradiation effects have been observed and are now under study. Microstructural changes in the fuel are dominated early in exposure by grain boundary cavitation and fission gas bubble growth, producing large amounts of swelling. Irradiation creep and swelling of the austenitic (D9) and martensitic (HT-9) candidate cladding alloys have been measured and correlate well with property modeling efforts. Chemical interaction between the fuel and cladding alloys has been characterized to assess the magnitude of cladding wastage during steady-state irradiation. Significant interdiffusion of the uranium and zirconium occurs producing metallurgically distinct zones in the fuel.

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This paper is based on a presentation made in the symposium “Irradiation-Enhanced Materials Science and Engineering” presented as part of the ASM INTERNATIONAL 75th Anniversary celebration at the 1988 World Materials Congress in Chicago, IL, September 25-29, 1988, under the auspices of the Nuclear Materials Committee of TMS-AIME and ASM-MSD.

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Pahl, R.G., Porter, D.L., Lahm, C.E. et al. Experimental studies of U-Pu-Zr fast reactor fuel pins in the experimental breeder reactor-ll. Metall Mater Trans A 21, 1863–1870 (1990). https://doi.org/10.1007/BF02647233

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Keywords

  • Metallurgical Transaction
  • Argonne National Laboratory
  • Metallic Fuel
  • Fuel Alloy
  • Fuel Slug