[en] Highlights: • Efficient recovery of actinides from irradiated metallic fuel METAPHIX-2 was demonstrated by electrorefining process. • High separation factors of actinides (An) over lanthanides (Ln) were achieved in electrolyte with high concentration of Ln. • Grouped selectivity of the actinides recovery was proven for later and final stages of the electrorefining process. • Aluminium cathode was proven suitable for recovery of all actinides, while only uranium was deposited on the inert cathodes. • Zirconium co-dissolution from the fuel had no effect on the process efficiency and on the structure of the deposits. - Abstract: An electrorefining process for homogeneous group-selective recovery of actinides from metallic nuclear fuel in molten LiCl–KCl has been investigated. The present study follows up on the previously achieved results on recovery of actinides from both non-irradiated and irradiated test metallic fuels. In the current work, METAPHIX-2 fuel initially composed of U₇₁–Pu₁₉-Zr₁₀ alloy irradiated to ∼7 at.% was processed. The experiments were focused on evaluation of selectivity of actinides over lanthanides during the electrorefining process in an electrolyte containing different concentrations of dissolved lanthanides up to 6.5 wt%, simulating the later and final stages of the process. In addition, a comparison of use of the solid reactive aluminium and solid inert cathodes for homogeneous recovery of all actinides was studied, as well as the effect of zirconium co-dissolution from the fuel on the process efficiency and on the structure of the deposits. The reactive aluminium electrode was proven suitable for homogeneous recovery of all actinides, while at the given conditions only uranium could be deposited on the inert cathodes. Very high group-selectivity of the process for actinides was demonstrated, even at high concentration of lanthanides in the electrolyte.

[en] Pyroprocessing technology, currently being developed in the Republic of Korea, transforms the spent fuels of a light water reactor into fresh MOX fuels for a sodium cooled fast reactor. It is important to manage nuclear material control in a pyroprocessing facility that treats nuclear materials such as U and Pu in view of safeguard. Until spent fuel assemblies transported from nuclear power plants are disassembled, and quantitative management of nuclear materials is carried out using a bulk item unit. However, quantitative management of nuclear materials is carried out in a unit of weight after the spent fuel assemblies are disassembled. Since the uncertainty that occurs when the quantitative management of nuclear materials is carried out in a unit of weight rather than by bulk item is relatively large, various methods are being developed for accurate quantitative management of nuclear materials to reduce the uncertainty of nuclear materials for safeguard measures.

[en] Nuclear fuel reprocessing is an important step in closing the nuclear fuel cycle. The most preferred method for the separation of actinides from the fission products is by solvent extraction. The versatile solvent tri-n-butyl phosphate (TBP) dissolved in a hydrocarbon diluent is commonly used for this purpose. Despite its excellent extraction behavior, TBP shows certain limitations, which affect its performance in fast-reactor fuel reprocessing. The drawbacks of TBP include its tendency to form third-phase in the extraction of tetravalent metal ions and chemical and radiation degradation. These limitations demand development of alternate extractants with improved performance, for the extraction of actinides with better properties than TBP. This requires molecular level understanding of the extracting solvent, metal ions and complex species formed during the extraction process. Structural parameters derived using spectroscopic and X-ray crystallographic techniques are highly useful and often compliment direct experimental observations on solvent extraction. Nevertheless, experimental ligand design for the separation of lanthanide and actinide metal ions from irradiated fuel is a tedious task

[en] The threshold-fission (fertile) nature of Am-241 is used to destroy this minor actinide by capitalizing upon neutron capture instead of fission within a sodium fast reactor. This neutron-capture and its subsequent decay chain leads to the breeding of even mass number plutonium isotopes. A slightly moderated target design is proposed for breeding plutonium in an axial blanket located above the active 'fast reactor' driver fuel region. A parametric study on the core height and fuel pin diameter-to-pitch ratio is used to explore the reactor and fuel cycle aspects of this design. This study resulted in both a non-flattened and a pancake core geometry. Both of these designs demonstrated a high capacity for removing americium from the fuel cycle. A reactivity coefficient analysis revealed that this heterogeneous design will have comparable safety aspects to a homogeneous reactor of the same size

[en] A 1000 MWth commercial-scale Sodium Fast Reactor (SFR) design with a conversion ratio (CR) of 0.50 was selected in this study to perform perturbations on the external feed coming from Light Water Reactor Spent Nuclear Fuel (LWR SNF) and separation groupings in the reprocessing scheme. A secondary SFR design with a higher conversion ratio (CR=0.75) was also analyzed as a possible alternative, although no perturbations were applied to this model

[en] Plasma methods for processing spent nuclear fuel are analyzed. It is shown that, by ICR heating in a nonuniform magnetic field, the energy of the heated ash ions can be increased substantially, while nuclear fuel ions can be kept cold. Two methods for extracting heated ash ions from a cold plasma flow are considered, specifically, that by increasing the ion gyroradius and that due to ion drift in a curved magnetic field. It is found that the required degree of separation of ash and fuel ions can be achieved in systems with quite moderate parameters

[en] The corrosion of the research reactor aluminium clad spent fuel in water was investigated in two storage facilities. The standard racks were delivered by the IAEA and consisted of two aluminium alloys AA 6061 and Szav-1 coupons. Bimetallic couples create aluminium alloy and stainless steel 304 coupons. Rolled and extruded AA 6061 material was also tested. Single coupons, bimetallic and crevice couples were exposed in the at-reactor basin (ARB) and the high-level wastage pool (HLW). The water chemistry parameters were monitored and sedimentation of impurities was measured. The content of impurities of mainly Cl and SO₄ was in the range of 2 to 15 μg/l in the HLW pool; it was about one order higher in ARB. The Fe content was below 2 μg/l for both facilities. After two years of exposure the pitting was evaluated as local corrosion damage. The occurrence of pits was evaluated predominantly on the surfaces of single coupons and on the outer and inner surfaces of bimetallic and crevices coupons. No correlation was found between the pitting initiation and the type of aluminium alloys and rolled and extruded materials. In bimetallic couples the presence of stainless coupons did not have any effect on local corrosion. The depth of pits was lower than 50 μm for considerable areas of coupons and should be compared with the results of other participating institutes. (author)

[en] After a briefly intraduction on the geological setting of basins in Honggeer area, this paper focuses on the sequence stratigraphy division of target prospecting Baiyanhua group of Lower Cretaceous and analysis on its sedimentary system, metallogenic prospect of sandstone type uranium deposit is dis- cussed in the view of uranium source, structure, sandbody scale and layer combination, oxidation and reduction, hydrogeology and uranium mineralization. Fovourable prospecting areas and further exploration deserving sections are proposed at the end. (authors)

[en] Undergoing sanctions due to the nuclear test explosions if conducted twice, and due to the fact that uranium is sold only to countries that are signatories to the NPT, India is facing a uranium crunch. The civil nuclear agreement between India and USA has to be viewed in this context. (author)

[en] Literature research was conducted on all kinds of nuclear fuel reprocessing technologies encompassing from commercial PREX process to developing new technologies in order to provide suitable information available for selection of reprocessing process technology applied to LWR-FBR transitional phase fuel cycle which will continue several decades from 2050. Development strategies for transitional phase fuel cycle were also investigated. Ideally, the spent MOX fuel from LWR should be reprocessed at the normal PUREX process and the technical problem must be enumerated and technical development must be started soon. Also, the spent MOX fuel from initial FBR may be treated at the PUREX process for LWR fuel by mixing the FBR fuel and LWR fuel, because not so much spent fuel will be discharged from the initial FBR reactors. If the spent fuels from LWR are treated at the reprocessing plant designed for FBR fuel, the U recovery technologies are necessary in order to reduce the mass flow at the extraction process. MA recovery may be required in future and new designed reprocessing plant must have the flexibility for appending new function such as MA recovery process or U and Pu purification process. (author)