[en] A kinetic-based model to predict the dissolution of UO₂ α-doped pellets under initial anoxic conditions is presented and compared with experimental results previously obtained. The uranium and plutonium concentrations in solution are predicted by considering the presence of a α-radiation field and its influence due to radiolysis of water on the pellet surface oxidation and subsequent dissolution. The initial parameters required by the model in order to reproduce the pellet alteration process are: system geometry, chemical composition of the leachant, physicochemical characteristics of the leachate and the oxidation conditions of the pellet surface (expressed in terms of U(VI)/U(IV) ratio). The last one is the key parameter in the model for simulating the initial quick dissolution process. The results obtained are compared with experimental data. The agreement between the predictions obtained and the published data is good. The influence on the matrix oxidation-dissolution process due to the α-radiation field intensity and the release of Pu are reproduced by the model. The Pu concentration trends as a function of time are explained in connection with the matrix dissolution process and are controlled by the formation of the secondary phase Pu(OH)₄(S)''. (Author)

[en] The formation of the high burnup structure (HBS) on the peripheral region of the light water reactor (LWR) UO₂ fuel pellet is characterized by a steep increase in porosity. The newly formed micron-size intergranular pores trap most of the created fission gas. The study of the pore size distribution and its evolution with burnup is fundamental to estimate the fission gas pressure in the pores and can influence the burst release in accidental conditions. Quantitative image analysis is usually employed to quantify the pore size-distribution. It is determined mainly in two dimension (2D) using a histogram. Few attempts have been made in the past to derive also three-dimensional (3D) pore size-distribution employing the Schwartz-Saltykov (ScSa) method. In this work, we determine experimentally the pore size-distribution in the HBS and apply a new procedure to determine the distribution, both in 2D and 3D. (authors)

No abstract available

[en] Metal fuel alloys containing 5 wt% or less minor actinide (MA) and rare earth (RE) were irradiated in the fast reactor Phénix. After nondestructive postirradiation tests, a chemical analysis of the alloys irradiated for 120 effective full power days was carried out by the inductively coupled plasma - mass spectrometry (ICP-MS) technique. From the analysis results, it was determined that the discharged burnups of U-19Pu-10Zr, U-19Pu-10Zr-2MA-2RE, and U-19Pu-10Zr-5MA were 2.17, 2.48, and 2.36 at.%, respectively. Actinide isotope ratio analyses before and after the irradiation experiment revealed that Pu, Am, and Cm nuclides added to U-Pu-Zr alloy and irradiated up to 2.0 - 2.5 at.% burnups in a fast reactor are transmuted properly as predicted by ORIGEN2 calculations. (author)

[en] The problem of irradiated fuel (both UO₂ and Mixed Oxide Fuels) interactions with liquefied Zircaloy at high temperatures is central to the understanding of bundle degradation mechanisms during reactor power transients or severe accidents. These initial interactions of the cladding and the irradiated fuel result in a melt (corium) and then to a loss of bundle geometry and the corium accumulation in a pool. ITU investigated the interaction of irradiated fuel and compared it with non-irradiated fuel with its Zircaloy cladding at 2000 deg. C for various short times. This was its contribution to the COLOSS (Core Loss of Geometry) project carried out under an EC framework programme. The tests were investigated by optical microscopy with image analysis and then by SEM-EDS analysis. The dissolution of the irradiated fuel by the Zircaloy melt was very variable and heterogeneous, but for non-irradiated fuel was reasonably uniform and constant. The kinetics of the non-irradiated UO₂-liquefied Zircaloy interactions was shown in another work package of the project to follow diffusion-limited mechanisms that could be modelled. The large variation in the results with the irradiated fuel rods made it difficult to model these interactions, nevertheless, they appear to have similar parabolic kinetics seen in non-irradiated fuel. The cracked condition of the fuel and the fission gas release during these interactions are major factors for fuel break-up, dispersion and dissolution in the melt under temperature transients.

[en] Leaching experiments at room temperature in demineralized water of UO₂ pellets doped with either ∼0.1 % and ∼10% wt. of the short-lived α-emitter ²³⁸Pu under anoxic conditions were performed. The results of sequential leaching tests showed significantly higher uranium dissolution for the material containing the ²³⁸Pu than for undoped UO₂. Essentially no α-emitter concentration effect was observed. The effect of different surface areas on the dissolution of UO₂ containing α-emitters was also investigated: preliminary results indicate similar radiolysis effects. Lattice parameter measurements as a function of storage time using XRD showed the expected rapid build-up of α-decay damage in the material with the higher concentration of ²³⁸Pu. A similar behavior was observed by measuring hardness as a function of time. During the same time intervals over a total of one year no clear variations of these parameters were detected for the material with ∼0.1 wt. % of additive

[en] Fast reactor metal fuels containing minor actinides (MAs) Np, Am and Cm and/or rare earths (REs) Y, Ce, Nd and Gd are being developed by CRIEPI in collaboration with JRC-ITU in the METAPHIX project.The test fuel pins including U-Pu-Zr-MAs(-REs) alloy rods in parts of the fuel stacks were fabricated and irradiated up to maximum burnup of ∼10 at.% in the fast reactor Phenix. At present the post-irradiation examinations (PIEs) are in progress at JRC-ITU. U-Pu-Zr-MAs(-REs) alloys of different composition were prepared for characterization experiments in advance of the test fuel pin fabrication. The miscibility of MAs in the U-19Pu- 10Zr(wt%)alloy was considered to be quite high: 15wt% Np and 10wt% Am could be mixed homogeneously in the U-Pu-Zr alloy. On the other hand, Am-REs-rich precipitates were formed in the alloys if REs were present. The metallographic observation of various U-Pu-Zr- MAs-REs alloys showed that such precipitates are dispersed homogeneously in the alloy provided that the in divisial amounts of MAs and REs added are limited to less than 5wt%. Furthermore, the physical properties such as elasticity, thermal conductivity, solidus temperature or phase transition temperature of U-Pu-Zr alloys are practically unchanged after the addition of ≤5wt% MAs and ≤5wt% REs. Based on the characterization results of the MA- and RE-containing alloys, standard U-19Pu- 10Zr fuel stacks including segments of U-19Pu-10Zr-2MA-2RE or U-19Pu-10Zr-5MA/U- 19Pu-10Zr-5MA-5RE alloy rods were fabricated for the irradiation experiment. A reference fuel pin of U-19Pu-10Zr alloy without MAs and REs was also fabricated and irradiated.The irradiation experiment was performed in the fast reactor Phenix (France) with the support of the Commissariat a l'Energie Atomique (CEA, France).The irradiated metal fuel pins were discharged from the reactor at ∼2.5at.%, ∼7at.% and ∼10at.% burnups in order to systematically examine fuel irradiation behavior and MAs transmutation rate. All the irradiations foreseen in this study were completed by May 2008. The non-destructive examinations of the irradiated metal fuel pins revealed that no damage due to the irradiation had occurred. The irradiated fuel pins up to ∼2.5at.% and ∼7at.% burnups were transported to JRC-ITU and are presently undergoing detailed destructive PIEs. The results of the measurement of the axial distribution of gamma-ray intensity for the ruthenium isotope ¹⁰⁶Ru, which hardly moves in the fuel alloy, showed fuel stack elongations of 1.9-2.5% (9-12mm) and 3.5-4.1% (17-20mm) in ∼2.5at.% and ∼7at.% burnup fuel pins, respectively. Fuel elongation behavior was independent of MAs and REs additions. Plenum gas analysis revealed that 45.7-51.3% and 63.1-68.2% of fission gases generated in the ∼2.5at.% and ∼7at.% burnup fuels, respectively, were released to the plenum; the total amount of fission gas generation was calculated using the ORIGEN-2 code. These results are consistent with reported data on the U-Pu-Zr test fuels irradiated in EBR-II as shown. The PIE results suggest that up to ∼7at.% burnup fuel swelling and fission gas release of U- Pu-Zr fuels containing ≤5wt% MAs and REs are essentially the same as those of MA- and RE-free U-Pu-Zr fuels

[en] In Switzerland, the spent nuclear fuel assemblies arising from the operation of the five NPPs are currently stored in pools at the NPP sites and, after a cooling period, are transferred to transport/storage casks which are then transported and stored in centralized dry interim storage facilities. The National Cooperative for the Disposal of Radioactive Waste (Nagra) has proposed deep geological disposal as the solution for the management of all radioactive waste. Pre-disposal activities, in particular for the spent fuel encapsulation facility and related unloading/loading and handling operations from the transport/storage casks into the final disposal canisters, are safety-relevant operations. Nagra therefore initiated several studies and RD&D activities aimed at assessing spent fuel mechanical performance, but also at developing concepts for handling of consequence scenarios. Concerning the RD&D program, the main objective of the investigations is to assess the response of spent fuel rods to mechanical stresses corresponding to normal conditions and accident scenarios by means of experiments on PWR spent fuel rod segments. The experimental campaign is conducted at JRC Karlsruhe, with the focus on the effect of hydrogen load, hydride distribution and pellet/cladding interaction on the cladding integrity. Other studies are currently under development to investigate the deterioration of the cladding properties resulting from Delayed Hydride Cracking (with Paul Scherrer Institute), as well as the deterioration of the FA structural material for long-term dry storage conditions (with Framatome GmbH). Furthermore, a conceptual study is under development to establish specific technical requirements for the encapsulation facility, focusing on fuel handling, retrieval and packaging operations. The main scope is to ensure the safe management of any damaged and degraded fuel and to implement measures for the mitigation of accident scenarios. Key aspects and main achievements of these ongoing programs are presented here. (author)

[en] The ternary oxide ceramic system UO₂-ZrO₂-FeO is a refractory system that is of great relevance to the nuclear industry as it represents one of the main systems resulting from the interaction of the Zircaloy cladding, the UO₂ fuel and the structural elements of a nuclear reactor. It is particularly the high temperature properties that require investigation; that is, when substantial overheating of the nuclear core occurs and interactions can lead to its degradation, melting and result in a severe nuclear accident. There has been much work on the UO₂-ZrO₂ system and also on the ternary system with FeO but there is still a need to examine 2 further aspects; firstly the effect of sub-oxidized systems, the UO₂-Zr and FeO-Zr systems, and secondly the effect of Fe/Zr or Fe/U ratios on the melting point of the U-Zr-Fe oxide system. Samples of UO₂-Zr and UO₂-ZrO₂-FeO were fabricated at ITU and then characterized by optical microscopy (OM) and X-ray diffraction to determine the ceramic's structure and verify the composition. Thereafter the samples are to be melted by laser flash heating and their liquidus and solidus temperatures determined by pyrometry. This programme is currently ongoing. The frozen samples, after testing, were then sectioned, polished and the molten zone micro-analytically examined by OM and SEM-EDS in order to determine its structure and composition and to compare with the existing phase diagrams. Examples of results from these systems will be given. Finally, a reacted Zr-FeO thermite mixture was examined, which had been used to generate high temperatures during tests of reactor melt-concrete interactions. The aim was to assess the reaction and estimate the heat generation from this novel technique. These results allow verification or improvement of the phase diagram and are of primary importance as input to models used to predict materials interactions in a severe nuclear accident.