[en] During steady-state irradiation of metal fuel in fast reactors, lanthanide fission products react with the Fe-base cladding and cause wastage of the cladding inner surface. In order to provide the basis of the cladding wastage modeling, the authors conducted isothermal annealing tests of diffusion couples consisting of Fe–12wt.%Cr alloy and lanthanide alloy, 13La–24Ce–12Pr–39Nd–12Sm (in wt.%), which simulates fission yield of lanthanide elements. In the temperature range of 773–923 K, Fe diffused into the lanthanide alloy side and formed Fe₂RE precipitates, where RE stands for lanthanide element mixture. Cr did not migrate evidently. The lanthanide elements diffused into the Fe–Cr side and formed the distinct reaction zone. This reaction zone showed two-phase structure of (Fe,Cr)₁₇RE₂ and (Fe,Cr)₃RE. Ce and Sm were concentrated in the Fe₂RE and (Fe,Cr)₁₇RE₂ phases. The thickness of reaction zone in the Fe–Cr side grew in proportion to the square root of annealing time. The activation energy of the reaction zone growth was determined, which can be the basis of the cladding wastage modeling

[en] Applicability of metallic U-Pu-Zr fuel to commercial fast reactors is being studied in Japan. This requires evaluation of the integrity of the metallic fuel to be used under severe irradiation conditions of commercial reactors, such as higher coolant temperature and higher burnup. When the high temperature application of the metallic fuel is considered, one of the fuel design limits is elimination of fuel centerline melting. The author presents review on some thermal properties of the fuel alloys for fuel temperature estimation. Another design limit is elimination of liquefaction at the fuel-cladding interface during normal reactor operation. The experiments and analyses to determine the liquefaction condition are summarized, where thermodynamic assessments of the relevant alloy systems play an important role as well as in estimation of the fuel alloy properties. The computer code prediction of the irradiation behavior up to a peak burnup ∼20at % is also presented. This suggests that appropriate estimations of cladding wastage by rare-earth elements and accumulation of non-gaseous fission products are important for the analysis of the metallic fuel performance up to such a high burnup. (author)

[en] For the fabrication of metallic fuel pins, injection casting is a preferable process because the simplicity of the process is suitable for remote operation. In this process, the molten metal in the crucible is injected into evacuated molds (suspended above the crucible) by pressurizing the casting furnace. Argonne National Laboratory has already adopted this process in the Integral Fast Reactor program. To obtain fuel pins with good quality, the casting parameters, such as the molten metal temperature, the magnitude of the pressure applied, the pressurizing rate, the cooling time, etc., must be optimized. Otherwise, bad-quality castings (short castings, rough surfaces, shrinkage cavities, mold fracture) may result. Therefore, it is very important in designing the casting equipment and optimizing the operation conditions to be able to predict the fluid and thermal behavior of the castings. This paper describes methods to simulate the heat and mass transfer in the molds and molten metallic fuel during injection casting. The results obtained by simulation are compared with experimental ones. Also, appropriate casting conditions for the uranium-plutonium-zirconium alloy are discussed based on the simulated results

[en] The quaternary U-Pu-Zr-Fe system was assessed using thermodynamic and phase diagram data in order to evaluate fuel-cladding chemical interactions (FCCI) of metallic fuel in a fast reactor. The Gibbs energy of mixing for solution phases and the Gibbs energy of formation of compounds in the binary sub-systems were calculated using an optimization procedure. The use of such data in optimizing the binary sub-systems enabled appropriate calculations for the thermodynamic properties of the systems, which were also important when extrapolating to higher-order systems. Isotherms of ternary sub-systems were calculated by using the optimized parameters of the binary sub-systems. Based on the phase relation data measured in regions of the ternary systems, the isotherms were then modified by adding ternary interaction parameters. The calculation results agreed well with the experimental data points. Finally, the quaternary system was assessed. The phase relationship observed experimentally in the diffusion couple of U-Pu-Zr-Fe was in reasonable agreement with the calculated phase diagrams. (authors)

[en] Gaps are formed between granular metal fuels and between metal fuels and a cladding tube, so that when metal fuels swell by radiation irradiation or generation of fission gases, the swelling is absorbed by reduction of the gaps between each of the metal fuels or movement of the metal fuels thereby preventing rupture of the cladding tube. Alternately, heat conductive mediums are packed between the metal fuels and between the metal fuels and a cladding tube to transfer the heat of metal fuels satisfactory to the cladding tube. Granular fuel alloys are charged into a cladding tube to pack a metal fuel, and in particular, the grain size of the metal fuels is sufficiently made smaller than the inner diameter of the cladding tube to further facilitate the packing. In addition, a mold is unnecessary in the granulation of the fuel alloys. (N.H.)

[en] Conclusion: Representative phases formed in FCCI were identified: • The reaction between lanthanide elements and cladding; • The reaction between U-PU-Zr and cladding (Fe). Characteristics of the wastage layer were clarified: • Time and temperature dependency of the growth ratio of the wastage layer formed by lanthanide elements; • Threshold temperature of the liquid phase formation in the reaction between U-Pu-Zr and Fe. These results are used: - as a basis for the FCCI modeling; - as a reference data in post-irradiation examination of irradiated metallic fuels

No abstract available

[en] An experimental study has been conducted to investigate the effect of freezing of liquid on the heat transfer characteristics for laminar flow between two cooled parallel plates of 20 mm in distance. Both plates were maintained at the same uniform temperature which was lower than the freezing temperature of the working fluid, water. The plate wall temperature ranged -2 ∼ -7 deg C and the inlet water temperature was varied from 1 to 8 deg C. The corresponding dimensionless wall temperature θ_w ranged 0.2 ∼ 4.5. The Reynolds number was varied from 700 to 2,300. The frozen layer was smooth and had a monotonously increasing thickness as water flowed downstream. In thin layer (θ_w ≤ 1.1), the experimental results agreed well with the analytical calculations. In thick layer (θ_w > 1.1), however, the measured layers were thinner than the calculated values since the fluid acceleration due to flow area reduction might produce an enhancement in heat transfer, which was neglected in the theory. (author)

[en] A model for constituent migration behavior in U-Pu-Zr metallic fast reactor fuel is proposed. It is based on diffusion equations for the ternary system under a radial temperature gradient, and it takes into account the alloy phase decomposition, assuming a quasi-binary U-Zr phase diagram with a constant plutonium content. Parametric simulations of Experimental Breeder Reactor II irradiation data with appropriate transport properties of the alloy system showed that the model can predict the experimentally observed radial three-zone structure and zirconium and uranium redistribution, although the predicted radial location of zirconium-depleted middle zone disagreed with the experimental result. Accumulation of basic experimental data on transport properties and a ternary phase diagram of the system are needed for a better understanding of the behavior

[en] The development of innovative nuclear fuel cycle technology, which has the advantages of economic and safe power generation as well as proliferation resistance, is strongly expected to be an effective measure in achieving environmental sustainability and satisfying the increasing energy demand. Metallic fuel cycle technology, consisting of a metallic fuel fast reactor, pyrometallurgical reprocessing and fuel fabrication by injection casting, is attracting increasing attention as one of the most promising nuclear fuel cycle technologies to achieve the above requirements. From the viewpoint of partitioning and transmutation, the metallic fuel cycle technology has excellent advantages. The metallic fuel which has a standard composition of U-Pu-Zr incorporates a very high content of the minor actinides (MA) homogeneously into itself and gives superior MA burning efficiency. In the pyrometallurgical reprocessing, no additional steps for MA recovery are needed, unlike for the aqueous processes since they always accompany the U-Pu product inevitably due to their similar thermodynamic properties. CRIEPI started research and development on the pyro-chemical processing and metallic fuel technologies in 1986 with domestic and international collaborations. Since 1994, CRIEPI and JAEA have jointly started studying the basics of actinide behaviours in molten salt and liquid metal systems, and expanded the joint study to carry out the integrated pyro-processing test and the metal fuel fabrication test for irradiation in the Joyo reactor. The basic feasibility of pyrometallurgical reprocessing, such as the recovery of uranium and transuranium elements by electrorefining, has already been confirmed. Since 2009, an engineering-scale fuel cycle test as a project entrusted by the Japanese Government has been in progress to obtain the data required for designing the pyro-processing equipment for practical use. The joint study between CRIEPI and JRC-ITU has demonstrated the irradiation integrity of MA containing metal fuels up to 10 at.pc BU and the recovery of MA from both irradiated metal fuels and spent MOX fuels. Regarding pyro-chemical technology, an additional advantage lies in its flexibility to accept other types of fuels, such as oxides and nitrides. Conventional UO₂ and MOX fuels can be supplied to the pyrometallurgical treatment after reduction to metals by adoption of the electrochemical reduction technique. In the joint study with JRC-ITU, it has been demonstrated that the irradiated MOX fuels can be successfully reduced to metals using this method. After this technological achievement and the adaptability to diverse materials of various physical/chemical properties, pyro-processing is currently under preliminary evaluation for its applicability to the treatment of corium, mainly consisting of (U, Zr)O₂, formed during the accident of the Fukushima Daiichi nuclear power plant. (authors)