Because minor actinide bearing fuel is still a new field, a tremendous amount of excellent research has been accomplished to date [7]. It appears that the greatest emphasis has been placed on heterogeneous target fuel concepts with a high content of minor actinides only (i. e., no plutonium). This trend may have been driven by the desire to set the performance envelope, namely to understand and prove performance characteristics in extreme cases. However, this approach may overlook some practical approaches amenable to commercial-scale deployment in practice.

The physicochemical properties of materials involved in the fuel design and fabrication of minor actinide fuels are needed to understand and improve performance of fuel elements exposed to high temperatures, cor­rosive coolants and a radioactive environment. The objectives for develop­ment of a physicochemical properties database are to assemble a database on fuel properties and performance sufficient to support a safety/licensing case, and to develop a fabrication and quality assurance process that will enable effective and economic fuel fabrication. The required physicochemi­cal properties include a range of thermodynamic (enthalpy and heat capac­ity, melting temperature, enthalpy of fusion, vapour pressure, vaporization, thermal expansion, density and surface tension), transport (thermal con­ductivity and diffusivity, emissivity and optical constants) and mechanical (hardness creep, thermal shock and swelling) properties.

These properties (Table 12.4) are needed to understand fuel performance and are also required for the modelling of fuel behaviour. Relevant experi­mental data for all minor actinide elements are not available for all proper­ties, and may include datasets for a mixture of actinides. The database is also sparse for metal, oxide, nitride and other potential fuel forms of the minor actinides. A database is available for a few Np compounds, while data for Am and Cm are sparse. The collection of property data for minor acti­nides is an ongoing effort.

The irradiation behaviour of minor actinide fuels may vary from that of conventional fuels in several ways. Most prominent is the increased fuel pin helium gas inventory due to capture and decay sequences associated with 241Am and a significant amount of 242Cm, which decays by a emission (half­life of 169 days) to 238Pu. The additional helium gas inventory can lead to higher fuel swelling rates and is an additional source term for fuel pin over­pressurization. An additional interesting phenomenon relates to the evolu­tion of isotopic mixtures in fuel with high 241Am and 237Np loadings, and the effect this evolution has on pin power. As plutonium isotopes are bred in fuel from neutron capture during irradiation, particularly in the thermal spectrum, fuel rod power increases as a function of irradiation time.

Other issues to be studied include fission product behaviour, and optimi­zation of the oxygen to metal ratio in oxide fuels. Due to the shift in isotopic composition of the starting fuel, the isotopic and chemical distribution of fission products also shifts relative to that of standard MOX fuel.

Many fuel types, encompassing different matrices and chemical forms, have been considered for minor actinide bearing fuels. Several unconven­tional fabrication techniques have been explored as well. The in-reactor irradiation tests conducted to date are all very promising. As post­irradiation examination results become available, they will provide valua­ble information to guide additional future irradiation tests. Obviously, more R&D and in-reactor irradiation tests are required to qualify minor actinide bearing fuels. The incorporation of MA has some impact on the physicochemical properties of fuel material. Some results are available for the incorporation of MA in MOX fuel (for example, lower melting tem­perature, influence of stoichiometry on thermal conductivity, redistribu­tion of Am). But additional data are needed to guarantee the safe operation of reactor and fuel cycle facilities (fuel fabrication and reprocessing).

In the US, AFC fuel test hardware was designed to simulate fast reactor test conditions in Idaho National Laboratory’s advanced test reactor ATR: the AFC-1 irradiation test series was designed to evaluate the feasibility of actinide-bearing fuel forms in sodium cooled fast reactors for the transmu­tation of actinides from spent nuclear fuel [8]. The fuel rods had the same diameter as EBR-II fuel, but were reduced in length. All fuels, both nitride and metal, were sodium bonded inside stainless steel Type 421 (HT-9) clad­ding with an inert plenum gas. AFC-1B, AFC-1F, and AFC-1^ irradiation test capsules provided irradiation performance data at intermediate burn ups of 4 to 8 at.% on non fertile and fertile actinide transmutation fuel forms containing plutonium, neptunium and americium isotopes. AFC-1D, AFC-1G, and AFC-1H capsules extended fuel performance with satisfac­tory results, in terms of gas release and microstructural data compared to U-xPu-10Zr fuel when correlated with fission density.

In Japan, the “Am-1” programme has been conducted in order to inves­tigate the irradiation behaviour of Americium containing MOX fuel in the experimental fast reactor Joyo [9, 10]. Two irradiation experiments were conducted in the Joyo MK-III 3rd operational cycle to research early thermal behaviour of MA-MOX fuel. Six prepared fuel pins included MOX fuel containing 3% or 5% americium (Am-MOX), MOX fuel containing 2% americium and 2% neptunium (Np/Am-MOX), and reference MOX fuel. Ceramography results showed that structural changes such as lenticu­lar pores and a central void occurred early, within the brief 10 min of irra­diation. The results of electron probe microanalysis revealed that the concentration of Am increased in the vicinity of the central void. Post irra­diation examination of these pins did confirm fuel melting and local con­centration evolutions under irradiation of NpO2-x or AmO2-x in the (U, Pu) O2-x fuel. These test results are expected to reduce uncertainties in the design margin for the design of MA-MOX fuels.

The SUPERFACT irradiation in Phenix (1986-1988) represents the main body of existing knowledge on in pile behaviour of MOX fuel loaded with MA [11]. This project demonstrated the feasibility of Am or Np incorpora­tion of up to 2% in MOX fuels. The main constraint in introducing minor actinides into the core (homogeneous recycling mode) is linked to their impact via core reactivity and kinetic factors.

For heterogeneous recycling and in particular in the case of blankets loaded with MAs, high MA content raises the question of managing the large quantity of He produced. Hence, a specific transmutation fuel micro­structure must be developed, which requires envisioning several innovative steps in irradiation systems.

Most fast reactor irradiation tests, like the SUPERFACT experiment, were done in the French Phenix SFR, which was finally shut down in 2009 [12]. Since there are only a handful of fast reactors still in operation that can provide prototypic irradiation test environments in fast spectrum, an international collaboration to expand test capabilities and optimize the limited availability of irradiation test facilities is desirable. For example, under the framework of the GEN-IV SFR programme [13], an international collaboration project called GACID (Global Actinide Cycle International Demonstration) is being conducted (2007-2016) with participation of French CEA, US DOE and Japanese JAEA. The objective of the project is to demonstrate the transmutation of minor actinides in a 20% Pu MOX fuel with the Monju SFR located in Tsuruga, Japan. The fuel pins will be manufactured at the French CEA Atalante hot cells in Marcoule, using USA MA feedstocks. Data obtained from the GACID irradiation project will provide a feasibility assessment of MOX fuel matrix for transmutation.

Partitioning and transmutation of radioactive waste 375