Franz Ingold

Zirconium nitride as inert matrix for fast systems

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.

Preparation of uranium nitride in the form of microspheres

Abstract Uranium nitride microspheres were fabricated by internal gelation and carbothermic reduction. The influence of the thermal treatment and the reaction atmosphere on the chemical composition and the structural parameter of the spheres were systematically investigated. High density (> 95% TD) spheres of 800 μm diameter were obtained by reacting in argon-hydrogen followed by nitrogen-hydrogen. Porous spheres with a distinct pore and grain structure and low crushing strength as feed for pellets have been fabricated in nitrogen and nitrogen-hydrogen atmosphere. Special emphasis was put on a reliable determination of nitrogen and on X-ray diffraction for the chemical composition and on the correlation of the crushing strength to structural parameters.

Cermet sphere-pac concept for inert matrix fuel

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.

X-ray diffraction analysis and data interpretation of stabilized zirconia inert matrix fuel doped with plutonium

Abstract Yttria stabilized zirconia doped with erbia and plutonia has been selected as an inert matrix fuel (IMF) at PSI in order to destroy fissile plutonium in the form of a uranium-free fuel in an effective way. The crystallographic structure (lattice parameters) of cubic zirconia strongly depends on the choice of the stabilizer and other dopants i.e. burnable poisons or fissile material. An extensive study of X-ray diffraction measurements was performed on zirconia samples containing different amounts of additives with the aim to observe lattice parameter and crystallite size changes in the IMF. A semi-quantitative model already available in literature was used and adapted to predict the “theoretical” lattice parameters of IMF with plutonia. The results show a good agreement of theory and experiment. Furthermore, for the first time the structure of active IMF based on zirconia has been investigated and been compared to the X-ray diffraction patterns of undoped zirconia. As a consequence, it is now possible to predict lattice parameters and final densities of IMF with varying compositions, and a good control of the sample dimensions during the fabrication can be guaranteed.

3.11 – Sphere-Pac and VIPAC Fuel

In todays applied light water reactor (LWR) technology, the fissile material is embedded in a ceramic matrix, pressed, and sintered to pellets, which are then filled into the cladding tube of fuel pins that are assembled to a fuel bundle. This is the most widespread and well-known concept, which is also mostly adapted for the present fast breeder reactor (FBR) technology. Many alternative fuel forms have, however, been researched, seeking simplified fabrication routes and other advantages. When fissile isotopes are coming from spent fuel that is chemically separated (reprocessed), particle fuel with its direct filling of fuel particles into the fuel pin offers several advantages. Two major types of particle fuel are discussed here: Sphere-pac and Vipac fuel.

Fuel fabrication processes, design and experimental conditions for the joint US-Swiss mixed carbide test in FFTF (AC-3 test)

Abstract The preparation of mixed carbide fuel for a joint (US-Swiss) irradiation test in the US Fast Flux Test Facility (FFTF) is described, together with the experiment design and the irradiation conditions. Two fabrication routes were compared. The US produced 66 fuel pins containing pellet fuel via the powder-pellet (dry) route, and the Swiss group produced 25 sphere pac pins of mixed carbide using the internal gelation (wet) route. Both sets of fuel met all the requirements of the specifications concerning stoichiometry, chemical composition and structure. The pin designs were as similar as possible. The test operated successfully in the FFTF for 620 effective full power days until October 1988 and reached over 8% burn up with peak powers of around 80 kW/m. The conclusions were that the choice of sphere pac or pellet fuel for reactor application is dependent on preferred differences in fabrication (e.g. economics and environmental factors) and not on differences in irradiation behaviour.

FUJI: A Comparative Irradiation Test with Pellet, Sphere-Pac, and Vipac Fuels

Abstract Particle fuels such as sphere-pac and vipac have been considered as promising fuel systems for fast reactors because of their inherent potential in remote operation, cost reduction, and incineration of minor actinides or low-decontaminated plutonium. The FUJI test addresses the questions of fabrication of mixed-oxide (MOX) particle fuels with high Pu content (20%) and its irradiation behavior during the start-up phase. Four kinds of fuel, i.e., MOX sphere-pac, MOX vipac, MOX pellet, and Np-MOX sphere-pac, have been and will be simultaneously irradiated under identical conditions in the High Flux Reactor in Petten, Netherlands. First results show that the particle fuel undergoes a substantial structure change already at the very beginning of the irradiation when the maximum power is reached. The changes in microstructure, i.e., the formation of a central void and the densification of fuel, decrease the fuel central temperature. Thus, the fast and strong restructuring helps to prevent central fuel melting at high power levels.

Design and fabrication aspects of a plutonium-incineration experiment using inert matrices in a “Once-Through-Then-Out” mode

Abstract In the plutonium incineration experiment, named ‘Once-Through-Then-Out’ (OTTO), that is being prepared by JAERI, PSI and NRG, the use of highly stable inert matrices will be examined. The inert matrices MgAl 2 O 4 spinel and ZrO 2 are insoluble in nitric acid and are considered as good storage media for final disposal. These inert matrices will be used in this experiment, which is representative for an OTTO scenario. A total of 7 Pu-containing targets were prepared for an irradiation in the High Flux Reactor in Petten. The objective of the irradiation is to reach a very high Pu-burnup. The main parameters to be studied are stability under irradiation, swelling, fission gas release and chemical interactions in the fuel. Four targets will be equipped with thermocouples for on-line monitoring of central temperature. Four of the targets contain MgAl 2 O 4 as an inert matrix, 2 targets contain ZrO 2 and one target contains mixed-oxide (MOX) fuel for reference purposes. The fissile plutonium concentration is 0.32–0.44 g cm −3 . Both particle-dispersed fuel and homogeneous dispersions were fabricated in order to test the effect of the size of the fissile inclusions. The design of the experiment and the fabrication of the samples are discussed.

Investigation of the UONa and (U,Pu)ONa phase diagrams — Study of Na3UO4 and Na3(U,Pu)O4 phases

Abstract The detailed study of UONa and (U,Pu)ONa phase diagrams, corresponding to Na 3 MO 4 (M = U or U,Pu) composition, by high temperature neutron diffraction, room temperature X-ray diffraction and microcalorimetry, has revealed the existence of metastable compounds and the formation of a solid solution by dissolution of MO 2 in Na 3 MO 4 . The existence of these new phases, with a disordered crystal structure, allows improved correlation with the previous results on the phase in equilibrium with MO 2 and Na.

Fabrication of neptunium containing MOX fuel microspheres

In the FUJI project (Fuel irradiation for JNC and PSI) Japan Nuclear Cycle Development Institute, NRG Petten and Paul Scherrer Institute will perform an irradiation experiment in the high flux reactor (HFR) at Petten. The behaviour of various types of fuels (MOX with 25 % Pu) under irradiation will be compared in order to give information on their suitability as fast reactor fuels. The present research is mainly focused on sphere-pac fuel containing high concentrations of plutonium and a significant amount of neptunium in the form of mixed oxide (MOX). This report describes the status of the preparation and characterization of optimized sphere-pac MOX fuel with 30 at.% Pu or 30 at.% Pu and 5 at.% Np, respectively. The main results and conclusions are the following: PSI has managed to produce single-phased cubic MOX fuel (with and without Np) for different sphere sizes and the desired compositions. It has been shown that the oxygen to metal ratio can be controlled not only for uranium oxide but also for MOX by selecting the proper sintering conditions. Some of the spheres show large pores, which have to be eliminated by optimizing the gelation procedure. Efforts to optimize the density of the material and to improve the analysis techniques (new thermogravimetric analysis equipment) are ongoing.