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Comprehensive nuclear materials 3 14 uranium intermetallic fuels (u–al, u–si, u–mo)Comprehensive nuclear materials 3 14 uranium intermetallic fuels (u–al, u–si, u–mo) Comprehensive nuclear materials 3 14 uranium intermetallic fuels (u–al, u–si, u–mo) Comprehensive nuclear materials 3 14 uranium intermetallic fuels (u–al, u–si, u–mo) Comprehensive nuclear materials 3 14 uranium intermetallic fuels (u–al, u–si, u–mo) Comprehensive nuclear materials 3 14 uranium intermetallic fuels (u–al, u–si, u–mo) Comprehensive nuclear materials 3 14 uranium intermetallic fuels (u–al, u–si, u–mo) Comprehensive nuclear materials 3 14 uranium intermetallic fuels (u–al, u–si, u–mo) Comprehensive nuclear materials 3 14 uranium intermetallic fuels (u–al, u–si, u–mo)
3.14 Uranium Intermetallic Fuels (U–Al, U–Si, U–Mo) Yeon Soo Kim Argonne National Laboratory, Argonne, IL, USA ß 2012 Elsevier Ltd All rights reserved 3.14.1 Introduction 392 3.14.1.1 3.14.1.2 3.14.1.3 3.14.2 3.14.2.1 3.14.2.2 3.14.2.3 3.14.2.3.1 3.14.2.3.2 3.14.2.4 3.14.2.4.1 3.14.2.4.2 3.14.2.4.3 3.14.2.5 3.14.3 3.14.3.1 3.14.3.2 3.14.3.3 3.14.3.4 3.14.3.4.1 3.14.3.4.2 3.14.3.4.3 3.14.3.5 3.14.4 3.14.4.1 3.14.4.2 3.14.4.3 3.14.4.3.1 3.14.4.3.2 3.14.4.4 3.14.4.4.1 3.14.4.4.2 3.14.4.4.3 3.14.4.5 3.14.5 References Background Historical Evolution of U Intermetallic Fuels Performance Topics of U Intermetallic Fuels U–Al U–Al Fuel Properties Thermal Conductivity of U–Al Alloy and UAlx–Al Dispersions U–Al Fabrication U–Al alloy UAlx U–Al Irradiation Performance Fuel swelling by fission products Interaction between U–Al and Al U–Al blister threshold temperature Summary for U–Al U–Si U–Si Fuel Properties Thermal Conductivity of (U–Si Intermetallic)–Al Dispersions U–Si Fabrication U–Si Irradiation Performance Fuel swelling by fission products Interaction between U–Si and Al U–Si blister threshold temperature Summary for U–Si U–Mo U–Mo Fuel Properties Thermal Conductivity of (U–Mo Alloy)–Al Dispersions U–Mo Fabrication U–Mo alloy powder fabrication U–Mo dispersion plate fabrication U–Mo Irradiation Performance Fuel swelling by fission products Interaction between fuel particles and Al matrix U–Mo alloy blister threshold temperature Summary for U–Mo Summary and Outlook 392 392 394 395 395 396 397 397 397 397 397 400 400 401 401 401 401 402 403 403 406 410 411 411 411 413 413 413 415 415 415 416 419 419 420 420 Abbreviations ANL ATR BU EFPD Argonne National Laboratory (Argonne, IL) Advanced test reactor (at INL) Burnup Effective full power days EOL EPMA ETR FD HEU End of life Electron probe microanalysis Engineering test reactor (at INL) Fission density in fuel phase High-enrichment uranium (usually $ 93 wt%235U) 391 392 Uranium Intermetallic Fuels (U–Al, U–Si, U–Mo) IL INL LEU MEU Interaction layer (reaction layer) Idaho National Laboratory (Idaho Falls, ID) Low-enrichment uranium ( Â 1027 fissions m3, DV %ị ẳ 3:0 ỵ 2:3fd 3ị ỵ 0:33fd 3ị2 V0 g ẵ23 (c) mm Figure 27 Scanning electron microscopy images of irradiated U–10Mo alloys The samples shown in this figure have the same fabrication history and similar irradiation temperatures (a) 35% 235U low-enrichment uranium (LEU) equivalent BU (b) 65% 235U LEU equivalent BU (c) 80%235U LEU equivalent BU higher burnup, bubbles uniformly span the entire fuel cross section (shown in Figure 27(c)) as the grain refinement is nearly completed Typically, total swelling of U–Mo fuel is obtained from plate thickness changes before and after irradiation The gas bubble swelling is obtained by subtracting the solid fission product swelling, given in eqn [4], from the measured total swelling The data are fitted to a linear function of fission density at low fission density and a parabolic function of fission density at higher burnup where fd is the fission density in 1027 fissions mÀ3 The measured fuel swelling data and prediction correlation composed of eqns [4], [22], and [23] are compared Figure 29 3.14.4.4.2 Interaction between fuel particles and Al matrix 3.14.4.4.2.1 Interaction in pure Al matrix IL formation between U–Mo fuel particles and matrix Al poses potential fuel failure risks because pores tend to form in thick ILs Figure 30 shows a cross section of postirradiation U–Mo dispersion in Al The dark gray circles are U–Mo fuel particles, the lighter gray areas around the fuel particles are the ILs, the lightest areas are the unreacted matrix Al, and the black shapes are pores formed in ILs The IL normally grows with smooth surfaces on the fuel particles The variable composition of the IL is possible because of its amorphous nature during irradiation Consistent results are available in the literature about amorphization of the IL.46,61 Amorphization of a crystalline material to metallic glass is usually accompanied by an increase in volume – a quantity Uranium Intermetallic Fuels (U–Al, U–Si, U–Mo) Table Differences of IL growth during irradiation from out-of-pile tests 80 Symbols: Measured data Lines: Prediction model 60 (ΔV/V0)f (%) 417 Temperature ( C) Time (h) Mechanism (ΔV/V0)s + (ΔV/V0)g 40 Recrystallization IL structure Out-of-pile In-reactor 500–580