[en] The FUJI project is especially dedicated to the early-in-life restructuring of two types of particle fuel 3/4 sphere-pac and vipac 3/4 in comparison to pellet fuel. Three types of sphere-pac fuel segments prepared by PSI will be irradiated in the High Flux Reactor (HFR) in Petten (NL), together with pellet-type fuel segments and vipac fuel segments (non-spherical, fuel-shard-packed segments), prepared in accordance with JNC specifications at PSI under the same irradiation conditions. MOX fuel, with 20% Pu and an oxygen-to-metal ratio of 1.97 (typical for fast reactor fuel), is currently being produced in different fuel shapes (pellets, microspheres and vipac particles). For the pellet and vipac fuel, the dry attrition mill method was used to prepare the MOX powder prior to pressing. Microspheres were produced following the internal gelation process, with nitrate solutions as the starting material. The pellet and the vipac fabrication have now been completed, although some analysis results are still missing. The special production route of vipac has led to particles of beneficial shape; the microsphere fabrication is underway. The preparation for the filling and welding of the segments have been completed, and production will take place before the middle of 2003. (author)

[en] The main objectives of the CONFIRM project of the European 5th Framework Programme are modelling, optimisation and fabrication of a uranium-free fuel for Accelerator-Driven Systems (ADS). Project partners are KTH (Stockholm), CEA (Cadarache), Studsvik Nuclear, BNFL (Risley), ITU (Karlsruhe) and SERCO (Harwell). In a first phase, plutonium fuel in a zirconium nitride matrix with high linear rating and high burnup will be irradiated in the Studsvik test reactor. The contribution of PSI is the production of (Pu,Zr)N pellets, the fabrication of four helium-bonded fuel rodlets, and their transport to Studsvik. Nitride pellets of low oxygen and carbon content, but with a rather high porosity, could be fabricated by means of the carbothermal reaction. (author)

[en] In the fabrication of fuel containing transuranium (TRU) elements, flow sheets and techniques that allow a shielded and/or remote fabrication will probably need to be applied. One approach, which has been demonstrated on the laboratory and semi prototype scale, is the wet fabrication route of co-precipitation of the matrix element uranium and plutonium to form either dense spherical particles or to produce hybrid pellets made from pressed gel microspheres. The ceramic material produced holds the TRU-elements (Pu, Np, Am) homogeneously distributed in the matrix. In conjunction with the Department d'Etudes des Combustibles of the French Commissariat a l'Energie Atomique (CEA-DEC) in Cadarache, PSI is further developing a mixed nitride ceramic and mixed oxide with high concentrations (up to 50%) of plutonium with the aim of a joint irradiation test of transuranium elements in the French PHENIX reactor. (author) 2 figs., 3 tabs., 16 refs

[en] A zirconia-based nuclear fuel for plutonium transmutation is being developed at PSI. Its composition would be ZrO₂-10%YO_1.5-7%ErO_1.5-15%PuO₂. The solubility and retention of cesium as well as formation of its zirconate need to be assessed, because cesium has been identified as a potential safety relevant fission product. Therefore, a critical study of the system Zr-Cs-O and an approach to describe the structural build-in of cesium ions in zirconia were performed. For comparison, the system U-Cs-O was selected in order to characterize the behaviour of cesium in uranium-free and uranium-based fuel. An extensive literature review was used to assess thermodynamic data for binary and ternary phases and yielded to a basis for calculating the phase relationships by using the software THERMO-CALC. The results are presented in form of isothermal sections (Zr-Cs-O and U-Cs-O at 1000 and 2000 K) and pseudobinary temperature/concentration sections (ZrO₂-Cs₂O, UO_2Cs₂O). In addition to the characterization of cubic stabilized zirconia, the formation of other ternary phases like orthorhombic Cs₂ZrO₃ and tetragonal Cs₂UO₄ was investigated by calculating ΔG_reaction-curves for different oxygen partial pressures. It has already been shown that zirconia can be stabilized by rare-earth oxides, and that their solubility as well as lattice parameters and ionic conductivity show a relationship with the dopant size. Further calculations to clarify the behaviour of alkali- and earth-alkali elements in zirconia are ongoing. (author) 4 refs

[en] The internal gelation process has traditionally been applied to fabricate standard fuel based on uranium, typically UO2 and MOX. To meet the recent aim to destroy plutonium in the most effective way, a uranium free fuel was evaluated. The fuel development programme at PSI has been redirected toward a fuel based on zirconium oxide or a mixture of zirconia and a conducting material to form ceramic/metal (CERMET) or ceramic/ceramic (CERCER) combinations. A feasibility study was carried out to demonstrate that microspheres based on zirconia and spinel can be fabricated with the required properties. The gelation parameters were investigated to optimise compositions of the starting solutions. Studies to fabricate a composite material (from zirconia and spinel) are ongoing. If the zirconia/spinel ratio is chosen appropriately, the low thermal conductivity of pure zirconia can be compensated by the higher thermal conductivity of spinel. Another solution to offset the low thermal conductivity of zirconia is the development of a CERMET, which consists of fine particles bearing plutonium in a cubic zirconia lattice dispersed in a metallic matrix. The fabrication of such a CERMET is also being studied. (author)

[en] Particle fuels, such as sphere-pac and vipac, are considered promising fuel systems for fast reactors, due for their inherent potential for remote, dust-free fabrication, the option for incineration of minor actinides, and the use of less costly, low-decontaminated plutonium. The FUJI test addresses the questions of fabrication of MOX particle fuels with high Pu-content (20%) and their irradiation behaviour in the reactor during the start-up phase. Four kinds of fuel, namely MOX sphere-pac, MOX vipac, MOX pellet and Np- MOX sphere-pac, have been, or will be, simultaneously irradiated under identical conditions in the High Flux Reactor in Petten. First results indicate that the particle fuel already undergoes a dramatic structural change at the very beginning of the irradiation, when maximum power is reached. The structural changes, i.e. the formation of a central void, and densification of the fuel, lead to a decrease of the fuel central temperature. Thus, the fast and intense restructuring helps prevent central fuel melting at high power levels. (author)

[en] At 16.15 on 24 May 1983 a rapid chemical reaction occured in an apparatus in which a routine waste sludge evaporation process was under way. The waste was produced during fuel fabrication. The resulting pressure wave ruptured a number of parts at the glove box in which the equipment was situated leading to an alpha contamination of that and neighboring laboratories. The alarm system activated by the fire alarm, the activity detectors and manualy by the operators functioned correctly. No one was physicaly injured and thanks to immediate protecture measures (e.g. gas masks) the amount of alpha-dust incorporated could be limited in degree and extent. Very time-consuming and costly cleaning and decontamination procedures were found to be necessary but these led to the encouraging result that the greater part of the contaminated equipment (particularly electrical and electronic apparatus) could be released for further unrestricted use in open labs. Detailed tests with the original waste solutions showed that at a certain ammonium nitrate content and at 235 C an exothermic chemical reaction can occur in the sludge. This temperature can only be reached due to a defective apparatus (e.g. a broken glass vessel) allowing the sludge to contact the heating element directly. The external expert agrees that this is the most likely cause of the incident. (author)

[en] In the inert matrix fuel concept, plutonium reprocessed from spent fuel is burned in an inert matrix, e.g. yttria-stabilized zirconia. Coming from wet reprocessing, the internal gelation can perform an easy micro-spheres production. Utilization of these particles in a sphere-pac realizes a direct fuel production. Besides being economical, this direct usage offers an almost dustless fabrication. One disadvantage of yttria-stabilized zirconia as matrix is its low thermal conductivity. A further reduction by the macroscopic structure of a sphere bed seems unacceptable. This can be eluded by the insertion of a highly conducting phase. Similar to the cermet concept with the embedment of ceramic fuel into metal, the infiltration of a fine metal fraction into a coarse ceramic fuel fraction is studied here. The initial thermal conductivity shows much higher calculated values and the sintering behaviour is also clearly enhanced compared to the pure ceramic bed

[en] The Internal Gelation Process, used for the fabrication of different nuclear fuel forms, offers for the nitrate-to-oxide conversion step in a mixed oxide fabrication flow sheet major advantages such as less mechanical process steps for a nearly dust free homogeneous product. For industrial application the utilization of microwaves for initiating the gelation of the formed drops offers a further potential in the reduction of process steps and makes the process even more suitable for remote operation. Sintered microspheres of three different size were produced and vibro-filled in 12 test rodlets, giving a fuel smear density in the rodlet of about 85% of the theoretical density. The rodlets were assembled into four strings (together with pellet and dummy rodlets) for irradiation in the Goesgen PWR in Switzerland. One rodlet string was withdrawn for detailed examination in each of the four years total irradiation time. The four cycle rod reached a burnup of >50 GW·d/t. The aim was to gain experience in sphere-pac uranium dioxide behaviour prior to testing uranium-plutonium oxides. In a similar way rodlets with MOX have been fabricated and are under irradiation in the Beznau plant in Switzerland. The Oxidative sintering of uranium oxide decreases the sintering temperature from >1700degC down to 1150-1200degC and shortens the sinter cycle period by a factor of 2. Microspheres of 500 micrometer diameter showed no significant difference in the compacting behaviour of uranium oxide converted by the AUC-process. The characterization of the sintered pellets indicates that the pellet densities sufficient for LWR application can be obtained with NIKUSI sintering. (author)

[en] Zirconium nitride has been proposed as inert matrix material to burn plutonium or to transmute long-lived actinides in accelerator-driven sub-critical systems or fast reactors. In combination with the possibility to fabricate specially shaped fuel pellets, e.g. barrel with a hole, an innovative fuel would be designed to reach higher burnup.The current project aims to use the direct coagulation casting process to shape mixed plutonium zirconium nitride annular pellets. As a backup option the conventional powder-pressing route is used to produce standard pellets. Gibbs energy calculations were used to optimise the thermal treatment of carborthermic reduction to yield mixed nitride powders. The status quo of the powder and pellet production as well as the first analytical results are presented in the following