G. Ledergerber

Preparation of simulated inert matrix fuel with different powders by dry milling method

Simulated zirconium oxide-based inert matrix fuel (IMF) has been prepared using cerium oxide replacing plutonium oxide, following the conventional pelletising route, in which an improved two-stage attrition mill was employed. Different methods have been applied for the powder preparation. In addition to powder mixing, nitrate solutions have been co-precipitated to bulk gel and to microspheres by the internal gelation process. The calcined porous microspheres were crushed in the attrition mill with a few passes. This co-precipitated powder was compared with three different mixtures of the four commercially available powders (ZrO2, Y2O3, Er2O3 and CeO2). Pellets of a standard size with relative densities higher than 90% TD and comparable grain and pore structures have been obtained. Based on these simulation tests both processes are found suitable for fabricating IMF pellets, which fulfil requirements of a commercial plutonium and uranium mixed oxide (MOX), where applicable. No intensive milling is required to form a solid solution in the sintering step for a (Zr,Y,Er,Ce)O2−x pellet. For the fabrication of plutonium-containing IMF, a similar behaviour can be expected.

Plutonium Incineration in LWR's by a Once-Through Cycle with a Rock-Like Fuel

Plutonium incineration in a uranium-free fuel by a once-through burning cycle in LWR’s followed by geological disposal of the rock-like material as a high level waste is discussed here. For burning plutonium of various origins, zirconium oxide is a promising candidate as inert matrix because it is stabilised by rare earth oxides (Er, Ho, Eu … Y) in a single phase solid solution with a stable cubic structure. In this material, selected rare earth isotopes can also act as burnable poisons. The spent fuel may be licensed as waste material on the basis of the inventory, the stability of the material and the behaviour of natural analogue material (e.g. baddeleyite). A fuel composed of 90-80% ZrO 2 , 7–14% PuO 2 and 3–6% Er 2 O 3 (At%), with potential addition of Y 2 O 3 (as additional stabiliser), is suggested for experimental study. Such a fuel employed in LWRs could generate power effectively while transmuting about 95% of the 239 Pu

Inert matrix fuel for the utilisation of plutonium

Utilising fuel resources responsibly, reducing waste volume and emissions as well as conflict potentials within the international community (non-proliferation, energy demand) are among the principles for the judgment of sustainable development. Utilising and burning plutonium in a light water reactor has been shown to be feasible for the disposition of the large amount of excess plutonium produced in todays power reactors and resulting from the disarmament efforts of the super powers. With regard to material technology aspects, efforts have concentrated on the evaluation of fabrication feasibility and on the determination of the physicochemical properties of a single phase zirconium/erbium/plutonium oxide material stabilised as a cubic solution by yttrium for plutonium and minor actinide incineration and transmutation. Due to the absence of uranium as origin for plutonium build-up, such a nuclear fuel is called Inert Matrix Fuel. Irradiation testing with a dedicated experiment in the material test reactor in Halden, Norway, is underway within the framework of the OECD-Halden Reactor Project.

Behaviour of implanted xenon in yttria-stabilised zirconia as inert matrix of a nuclear fuel

A zirconia-based fuel is studied for use of plutonium in light water reactors. Among the relevant properties for a nuclear fuel, efficient retention of fission products is required since the fuel matrix constitutes the first barrier against fission product release. To study the retention of xenon, its stopping power and its diffusion properties within (Er,Y,Pu,Zr)O2 potential inert matrix fuel (IMF) are investigated. Stopping and range of ions in matter (SRIM) calculations were carried out to estimate the average penetration depth of Xe ions as a function of their incident energy and of the material composition. To study its diffusion properties, Xe was implanted into yttria-stabilised zirconia (YSZ) to a depth of around 100 nm from the surface. After successive heat treatments to a maximum temperature of 1773 K, quantitative Xe depth profiles were determined by Rutherford backscattering. No profile modification by diffusion was observed. The behaviour of Xe is investigated at the subnanoscopic level and compared with the results obtained with zirconia samples implanted with Cs or I, as well as with Xe in UO2.

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.

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.

Design study of an irradiation experiment with inert matrix and mixed-oxide fuel at the Halden boiling water reactor

Abstract An effective way to reduce the large quantities of plutonium currently accumulated worldwide would be to use uranium-free fuel in light water reactors (LWRs) so that no new plutonium is produced. To test such a new fuel under reactor conditions and in comparison with standard mixed-oxide (MOX) fuel, an irradiation experiment is planned at the Halden boiling water reactor. The behaviour of three fuel rods consisting of uranium-free fuel will be investigated together with three rods made out of uranium–plutonium mixed-oxide fuel in the same assembly. The fuel compositions were adjusted so that all rods produce a similar power. Because of the moderation with D 2 O in the Halden reactor, two different surroundings of the considered assembly were examined to analyze the influence of the flux spectrum on the experiment. This showed that the influence of the spectrum on the material behaviour is negligible. The relation between assembly power and average neutron detector signal as well as the burnup or depletion function was calculated. The assumed power history was adapted to a usual LWR schedule. It is possible to reach a burnup of ∼540 MW d kg HM −1 with the uranium-free fuel and ∼54 MW d kg HM −1 with the MOX fuel after five years of irradiation, which is similar to the average burnup reached in commercial LWRs after four years of operation.

Measurement of the mechanical properties of thermally-shocked zirconia-based simulated inert matrix fuel

The Vickers micro-hardness (H V ) was measured by an indentation technique of simulated ZrO 2 -based Inert Matrix Fuel (IMF) material with a composition of Er 0.07 Y 0.10 Ce 0.15 Zr 0.68 O 1.915 in two different densities on sintered specimens and specimens thermally shocked with the quenching temperature differences (ATs) between 473 and 1673 K and compared with those of simulated MOX, namely, U 0.92 Ce 0.08 O 2 . The H V values obtained for two IMF materials were found higher, ranging from 6.37 GPa to about 7.84 GPa, depending on AT and the sintered density, than those obtained for the simulated MOX which are quasi-constant in the same range of AT with a mean value of 6.37 GPa. The fracture toughness (K IC ) was calculated from the measured H V and the crack length, and it was found to exhibit a slight increase with increasing AT, ranging between 1.4 and 2.0 MPa m 1/2 , while that of simulated MOX specimen is ranging between 0.8 and 1.1 MPa m 1/2 . The thermally shocked specimens were observed with an optical microscope and analyzed in terms of microstructural changes and cracking patterns.