Outline of the METAPHIX program and the post-irradiation examination data described below are published in Refs. [1-10],
METAPHIX is the collaborative program between CRIEPI and the European Commission Joint Research Centre - Karlsruhe (JRC-Karlsruhe) being conducted to understand the irradiation behavior of fast reactor metal fuel containing minor actinides (MAs: Np, Am, Cm) and demonstrate the MA transmutation performance in fast reactors. In this program, metallic fuel pins that contain U-Pu-Zr-MA alloys were fabricated at JRC - Institute for Transuranium Elements (JRC-ITU, currently JRC Karlsruhe) and irradiated in the Phénix fast reactor up to three different burnups, approximately 2.5at.% (METAPHIX-1), 7at.% (METAPHIX-2) and 10at.% (METAPHIX-3), with the support of the Commissariat a l´Energie Atomique et aux Energies Alternatives (CEA, France).
Since the MAs to be recovered pyrometallurgically from spent fuels entrain comparable amounts of rare earth fission products due to their chemical affinities, the effect of the contamination of MAs with REs should be examined. The characterization of U-Pu-Zr alloys containing MAs and rare earth elements (REs) indicated that practically homogeneous fuel alloys can be prepared provided that the total content of REs is limited to 5 wt.% or less. Based on this characterization result, three MA-bearing fuel alloys, U-19Pu-10Zr-2MA-2RE, U-19Pu-10Zr-5MA-5RE and U-19Pu-10Zr-5MA (wt.%), were selected along with a standard ternary fuel alloy, U-19Pu-10Zr, for the irradiation experiment. In the METAPHIX program, Np, Am and Cm were used as MAs, and Y, Ce, Nd and Gd were used as REs.
Three metal fuel pins, two of which include MA-containing alloy segments of 100mm length in the U-19Pu-10Zr fuel stacks, were fabricated. The total length of the active fuel was 485mm. The cladding material was 15-15Ti austenitic steel and the fuel-cladding gap was filled with sodium for thermal bonding.
The specification of METAPHIX-2#1 fuel pin is identical to that of METAPHIX-1#1 fuel pin.
References
[1] Paxton, M.M., et al., Comparison of the in-reactor creep of selected ferritic, solid solution strengthend, and precipitation hardened commercial alloys. Journal of Nuclear Materials, 1979. 80(1): p. 144-151.
[2] Paxton, M.M., B.A. Chin, and E.R. Gilbert, The in-reactor creep of selected ferritic, solid solution strengthened, and precipitation hardened alloys. Journal of Nuclear Materials, 1980. 95(1–2): p. 185-192.
[3] Straalsund, J.L., R.W. Powell, and B.A. Chin, An overview of neutron irradiation effects in LMFBR materials. Journal of Nuclear Materials, 1982. 108–109(0): p. 299-305.
[4] Straalsund, J.L. and D.S. Gelles, Assessment of the Performance of the Martensitic alloy HT-9 for Fast Breeder Reactor Applications. Topical Conference On Ferritic Alloys For Use In Nuclear Energy Technologies, 1983.
[5] Puigh, R.J. and G.L. Wire, In-reactor Creep Behavior of Selected Ferritic Alloys. Preceedings of Topical Conference on Ferritic Alloys for Use in Nuclear Energy Technologies, 1984.
[6] Chin, B.A., Analysis of the Creep Properties of a 12Cr-1 Mo-W-V Steel. 1984: p. 593-599.
[7] Huet, J.-J., et al., Swelling of ferritic steels irradiated in fast reactors. Irradiation Behavior of Metallic Materials for Fast Reactor Core Components, 1979: p. 5-9.
[8] Puigh, R.J. and F.A. Garner, Irradiation Creep Behavior of the Fusion Heats of HT-9 and Modified 9Cr-1Mo. Effects of Radiation Materials: Fourteenth International Symposium.
[9] Garner, F.A. and R.J. Puigh, Irradiation creep and swelling of the fusion heats of PCA, HT9 and 9Cr-1Mo irradiated to high neutron fluence. Journal of Nuclear Materials, 1991. 179–181, Part 1(0): p. 577-580.
[10] Toloczko, M.B., et al. Variability of Irradiation Creep and Swelling of HT9 irradiated to high Neutron Fluence at∼400°C to 600°C, Effects of Radiation Materials: Eighteenth International Symposium 1999, 765.
Dataset provided by CRIEPI (Japan) and the European Commission Joint Research Centre - Karlsruhe (JRC-Karlsruhe) in the framework of IAEA CRP T12031” FUEL MATERIALS FOR FAST REACTORS (FMFR) (2019-2023)”.
The data and accompanying documentation are supplied with the understanding that any utilization of these resources sourced from the IAEA fuel database or locally adapted versions thereof, leading to a publication (be it a journal article, conference proceeding, laboratory report, book, etc.), requires the acknowledgment in said publication of both the IAEA fuel database and the data’s respective authors or originating laboratories.
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