A. Leenaers

From High to Low Enriched Uranium Fuel in Research Reactors

Since the 1970s, global efforts have been going on to replace the high-enriched (>90% 235U), low-density UAlx research reactor fuel with high-density, low enriched (<20% 235U) replacements. This search is driven by the attempt to reduce the civil use of high-enriched material because of proliferation risks and terrorist threats. American initiatives, such as the Global Threat Reduction Initiative (GTRI) and the Reduced Enrichment for Research and Test Reactors (RERTR) program have triggered the development of reliable low-enriched fuel types for these reactors, which can replace the high enriched ones without loss of performance. Most success has presently been obtained with U3Si2 dispersion fuel, which is currently used in many research reactors in the world. However, efforts to search for a replacement with even higher density, which will also allow the conversion of some high flux research reactors that currently cannot change to U3Si2 (eg. BR2 in Belgium), have continued and are for the moment mainly directed towards the U(Mo) alloy fuel (7-10 w% Mo). This paper provides an overview of the past efforts and presents the current status of the U(Mo) development.

On the solubility of chromium sesquioxide in uranium dioxide fuel

Chromium sesquioxide doped uranium dioxide systems with a variety of initially added Cr2O3 amounts were prepared under different sinteringconditions. The solubility limit of Cr for each sinteringcondition is derived from the measured content of the dopant dissolved in the UO2 matrix usingelectron probe microanalysis (EPMA). It is found that the solubility of chromium in a uranium dioxide system for these series sintered at 1600, 1660 and 1760 C is limited to respectively 0:065 � 0:002, 0:086 � 0:003 and 0:102 � 0:004 wt% Cr. The lattice parameters of the different Cr2O3 doped fuels have been examined with X-ray diffraction (XRD). The XRD peaks of the samples sintered at 1760 Ca s well as a UO2 reference without Cr, prepared under the same conditions, were measured and a value for the lattice parameter a for each sample was obtained usingthe unit cell refinement method. A slig ht contraction of the lattice parameter is observed with increasingdopant content. 2003 Elsevier Science B.V. All rights reserved.

Characterization of Uranium Particles Produced by Hydrolysis of UF6 Using SEM and SIMS

Environmental sampling (ES) is a powerful technique used by safeguards inspectors of the International Atomic Energy Agency and the European Safeguards Office for the detection of undeclared nuclear activities. Since its implementation in the 1990s, ES has proven to be very sensitive and effective. Considering the consequences, the measurements should be carried out under a quality management programme. At the Institute for Reference Materials and Measurements, a new production method is under development for the preparation of reference uranium particles from well-certified UF6, allowing uranium particles with certified isotopic abundances to be prepared that are representative of those found in uranium enrichment facilities. Using an aerosol deposition chamber designed and built for the purpose, particles are formed by the hydrolysis of UF6 and their morphology and (isotopic) composition measured using SEM-EDX and SIMS. The SEM measurements show that by varying the relative humidity of the air in the reaction chamber, the morphology of the particles can be changed. By making a distribution map of the chemical composition of the particles, the relationship between fluorine and uranium as main constituents of the particle could be established. The presence of fluorine is a valuable indicator for the occurrence of nondeclared enrichment activities.

Oxidation of spent UO2 fuel stored in moist environment

Spent fuel remnants, cut from BWR fuel rods were retrieved for microscopic investigations after 25 years of storage at ambient temperature. The storage atmosphere of the samples studied here was dry air for 10 years and a moist environment for the remaining 15 years. The comparison with earlier experiments conducted on samples that were stored in dry air for the full period demonstrates that moisture accelerates the oxidation process also at the low temperatures considered here. The composition of the atmosphere in contact with the fuel has been analysed by mass spectrometry and the microstructure of the samples has been investigated by optical and scanning electron microscopy. The microscopic observations evidence that grain boundaries are affected and different stages in grain boundary alteration could be evidenced. Oxidation first induces a weakened intergranular bond followed by increased sensitisation to chemical attack and finally decohesion and onset of bulk oxidation.

Bituminised Waste Re-treatment: Replacement of the Bitumen Matrix by a Glass Matrix

The vitrification of nuclear waste salts similar to the salts present in Eurobitum produced by the former Eurochemic reprocessing plant, Mol-Dessel, Belgium, is investigated. In order to approach the composition of the high-level waste reference glass SON68, mixtures with different ratios of glass frit to waste salts have been prepared and vitrified. The concentration of the waste salts varied between 17 and 25 % by weight. To compare the short term corrosion behaviour with reference waste glass, leach tests have been performed. During 8 weeks, the glass samples were submitted to leach tests at 363 K using Milli Q water as leachant, at a glass exposed area over leachant volume ratio equal to 100 m −1 . During this short period, scanning electron microscopy with energy dispersive X-ray analysis (SEM-EDX) showed no gel layer formation. This technique did not detect changes in the chemical composition. However, X-ray photoelectron spectroscopy (XPS) showed that the binding energy of the Zr 3d and Si 2p electrons decreased as a function of leaching time, indicating an increasing covalent character of the glass surface. Inductively coupled plasma-mass spectrometry (ICP-MS) of the leachate as a function of time shows that ionic species such as Na and K leach faster than e.g. Si and B. These results together with the XPS data suggest the onset of a gel layer formation. The effect of waste content in the glass on the leaching rate is also studied. The results are compared with similar leaching tests on R7T7 type SON68 glass. Most elements are better retained in the simulated Eurobitum glass than in the SON68 glass.

Anisotropic thermal expansion of Ni, Pd and Pt germanides and silicides

Silicon or germanium-based transistors are nowadays used in direct contact with silicide or germanide crystalline alloys for semiconductor device applications. Since these compounds are formed at elevated temperatures, accurate knowledge of the thermal expansion of both substrate and the contact is important to address temperature depending effects such as thermal stress. Here we report the linear coefficients of thermal expansion of Ni-, Pd- and Pt-based mono-germanides, mono-silicides and di-metal-silicides as determined by powder-based x-ray diffraction between 300 and 1225 K. The investigated mono-metallic compounds, all sharing the MnP crystal structure, as well as Pd2Si and Pt2Si exhibit anisotropic expansion. By consequence, this anisotropic behaviour should be taken into account for evaluating the crystal units cell at elevated temperatures.

Status of the Low Enriched Uranium Fuel Development for High Performance Research Reactors

Historically, uranium enriched to >90% 235U has been used for many peaceful applications requiring high fission densities such as driver fuels for research reactors. However, the use of high-enriched uranium or HEU (all enrichments >20% 235U are considered HEU) for civil applications, is considered a proliferation concern. Since the 1970s, efforts are being devoted to the conversion of research reactors operating on HEU to alternative fuels using uranium with enrichment below 20% or LEU. These efforts imply the development of high-density LEU fuels to replace the low volume-density (mostly) UAlx based HEU fuels. The paper updates the present status of these developments focusing on the UMo dispersion fuel. It aims to provide an overview of the knowledge generated and the lessons learned in roughly 15 years of UMo dispersion fuel R&D in Europe through irradiation experiments and post-irradiation examinations (PIE).