Nuclear fuel devices of Pressurised Water Reactors are composed of uranium oxide pellets which are enclosed in zircaloy cylinders. During reactor operation different processes occur. In the contact with the fuel the zircaloy oxidises nonuniformely in depths of order of some micrometers. Further on, energy deposition of fission recoils leads to sputtering of uranium onto the inner surface of the cladding material. Thus, sputtered uranium ions start to migrate outwards. This paper presents first the results of experiments performed on the Lohengrin spectrometer in order to simulate the alteration of cladding tubes in contact with the nuclear fuel during irradiation. The energy loss of selected fission products is correlated to an oxygen mass gain. From these experiments, we deduce the oxidation kinetics constant under irradiation at a mean temperature of 400°C. These results are compared with those of a thermal oxidation. Following oxidation the fission product kinetic energy is stabilised, but still a broadening of the energy distribution is observed, which is characteristic of actinide diffusion inside the zirconia target. Using the Fick model, an actinide diffusion coefficient into zirconia under irradiation is deduced.

The trapping mechanism of hydrogen in niobium has been investigated within the molec­ ular dynamics approach. Simulations of the diffusion process of this impurity were performed which require the knowledge of the interatomic potential. N b N b interaction was described by an N body potential. The potential parameters were adjusted with respect to static and dynamic properties of Nb crystal. Nb H interaction was represented by a two body poten­ tial. The Arrhenius diagram of the H diffusion coefficient achieved by molecular dynamics in the single crystal case provides too small an activation energy in comparison with exper­ imental results. However, molecular dynamics simulations indicate a large increase of these values in the presence of defects. It is only at around 1000 K that the diffusion of hydrogen is not altered by defects. This conclusion confirms the experimental results concerning a good characteristic of superconducting cavities after thermal treatments. t Soumis a : Phys. Rev. C Molecular dynamics simulation of hydrogen diffusion in niobium. Influence of point defects \ ) B. Roux, H. Jafi"rezic, A. Chevarier, N. Chevarier InlJtitut de Physique Nucleaire de Lyon, IN2P3GNRS et Universite Glaude Bernard 43, Bd du 11 Novembre 1918, 69622 Villeurbanne Gedez, France PACS nurnber(s) : 61.72 Ss, 66.30 Jt

In the context of nuclear waste management, the release of hydrogen can generate safety problems. It has been shown that amorphous cobalt hydroxo-sulphide is able to decrease the apparent production of radiolytic hydrogen yield by hydrocarbon radiolysis. This paper presents a study of the effect of cobalt hydroxo-sulfide on the radiolysis of propane induced by a proton beam. Hydrogen and hydrocarbon partial pressures obtained after propane irradiation were compared in different conditions: (i) without cobalt hydroxo-sulphide, (ii) in the presence of solid, irradiated or not, by adjusting the gas pressure in the irradiation cell. The evolution of those partial pressures as a function of the time of contact between the amorphous cobalt hydroxo-sulphide and the formed gaseous species is discussed. The solid is characterised by X-ray diffraction, infra-red and Raman spectroscopies before and after hydrogen trapping. It is shown that the solid does not change the composition and amount of organic radiolytic products but acts as a trap of hydrogen. The hydrogen consumption by the solid seems to be proportional to the initial hydrogen partial pressure. This study shows that radiolytic hydrogen is rapidly produced by propane irradiation whereas the trapping of hydrogen by cobalt hydroxo-sulphide is a slower phenomenon.

The objective of the present paper is to put in evidence the influence of damages due to the collision cascades which take place at the end of the heavy ion micrometer range, on the zirconium surface oxidation. A comparison between two zirconium oxidation experiments under heavy ion irradiation performed in the same temperature and pressure conditions is presented. In the first experiment, a 6 mu m thick zirconium foil is irradiated with a 50 MeV Xe-129 beam which stops in the Zr foil. In the second one, a 2 mu m thick zirconium foil is crossed by the whole fission fragment distribution emitted in the U-235 fission. In order to explain the unexpected increase of the oxidation rate observed in the first case, MeV argon irradiation experiments have been performed at IPNL. Results show that collision cascades created in the Zr foil, far from the surface, increase significantly the Zr oxidation. (c) 2006 Elsevier B.V. All rights reserved.

Zirconolite is a candidate host material for conditioning minor tri- and tetra-valent actinides arising from enhanced nuclear spent fuel reprocessing and partitioning, which can be disposed in a geological repository for nuclear waste. Its chemical durability has been studied here under charged particle-induced radiolysis (He2+ and proton external beams) to identify possible effects on dissolution rates and mechanisms in pure water. Two geometries of experiments have been used to evaluate the influence of the following parameters: solid irradiation, Linear Energy Transfer (LET) at the interface and total deposited energy. Preliminary results on the elemental releases due to the enhanced dissolution of the zirconolite surface during charged particle-induced irradiation are first presented. Then, we focus on H2O2 production which is one of the major molecular species, created under water radiolysis, and likely to interact with the zirconolite surface. In presence of zirconolite, first results indicate an apparent consumption of the radiolytic hydrogen peroxide or its precursors compared to the production in pure water calculated from the primary yield GH(2)O(2). The measured H2O2 concentration varies linearly with the total deposited energy in water over the irradiation duration (between 1 h and 6 h) and in the conditions of our experiments. Moreover, the H2O2 concentration decreases when the local density of the deposited energy close to the interface increases. Thus, we suggest that the mechanism(s) leading to the consumption of H2O2 or its precursors involve zirconolite surface reactions.

This paper concerns the study of zirconium oxidation under irradiation with high energetic Xe ions. The irradiations were performed on the IRRadiation SUD (IRRSUD) beam line at Grand Accélérateur National d’Ions Lourds of Caen. The oxygen partial pressure was fixed at 10−3Pa and two temperature conditions were used, either 480°C reached by Joule effect heating or 280°C due to Xe energy deposition. Zirconia was fully characterized by Rutherford backscattering spectrometry, scanning electron microscopy, and grazing angle x-ray diffraction. Apparent diffusion coefficients of oxygen in ZrO2 were determined from these experiments by using a model which takes into account a surface exchange between oxygen gas and the ZrO2 surface. These results are compared with thermal oxidation data.

Nuclear fuel devices of Pressurised Water Reactors are composed of uranium oxide pellets which are enclosed in zircaloy cylinders. During reactor operation, major actinides are created among which plutonium 239. Further on, energy deposition of fission recoils leads to sputtering of uranium and plutonium onto the inner surface of the cladding material. Thus, sputtered Pu and U ions start to migrate outwards. Experiments to investigate the migration process were performed at the Institut Laue Langevin (Grenoble) on the Lohengrin mass spectrometer. A model is proposed to deduce an apparent uranium and plutonium diffusion coefficient in zirconia (ZrO2) from the energy distribution broadening of a selected fission product.

This paper deals with the comparison between thermal and Fission Enhanced Diffusion (FED) of uranium into zirconia, representative of the inner face of cladding tubes. The experiments under irradiation are performed at the Institut Laue Langevin (ILL) in Grenoble using the Lohengrin spectrometer. A thin 235UO2 layer in direct contact with an oxidised zirconium foil is irradiated in the ILL high flux reactor. The fission product flux is about 1011 ions cm−2 s−1 and the target temperature is measured by an IR pyrometer. A model is proposed to deduce an apparent uranium diffusion coefficient in zirconia from the energy distribution broadening of two selected fission products. It is found to be equal to 10−15 cm2 s−1 at 480 °C and compared to uranium thermal diffusion data in ZrO2 in the same pressure and temperature conditions. The FED results are analysed in comparison with literature data.

This paper presents a fundamental study of the radiolysis of gaseous organic molecules induced by proton beam. For that purpose, a specific extracted beam line associated with a gas irradiation cell was set up on the 4 MV facility of the Institut de Physique Nucleaire of Lyon. The first experiments have been performed with gaseous alkanes and alkenes. The gaseous species formed during irradiation are analysed by an on-line gas chromatography instrument equipped with two detectors. In order to test our experimental facility, we have studied the influence of irradiation parameters (duration, beam intensity, pressure) on the production of hydrogen. In the case of propane, the radiolytic yield value of hydrogen G(H-2) is equal to 3.7 for total doses in the range of 0.4-2.3 MGy at atmospheric pressure. (c) 2005 Elsevier B.V. All rights reserved.

This paper is devoted to the study of the defect influence on uranium diffusion in zirconia in the context of nuclear waste disposal. The experiments in reactor conditions are performed at the Institut Laue Langevin in Grenoble using the Lohengrin spectrometer. A thin UO2 layer in direct contact with a zirconium foil is irradiated in the ILL high-flux reactor. The fission product rate is around 3 x 10(11) ions s(-1) and the neutron flux is equal to 5 x 10(14) n cm(-2) s(-1). The target temperature is controlled by an IR pyrometer and ranges from 470 to 490 degrees C. In these conditions, a zirconium oxidation is first observed before uranium diffusion. A model is proposed to deduce an apparent uranium diffusion coefficient in zirconia (ZrO2) from the energy distribution broadening of a selected fission product (A=90). It is found to be equal to 10-(15) cm(2) s(-1).The study of thermal diffusion is performed by using ion beam techniques (ion implantation and Rutherford Backscattering). ZrO2 samples are implanted with 800 keV uranium ions at a dose of 10(16) ions cm(-2) and annealed at a pressure of 7.5 X 10(-1) Pa. No uranium diffusion could be observed up to 800 degrees C.The influence of irradiation defects mainly due to fission products, both on zirconium oxidation and on uranium diffusion, is clearly demonstrated. (c) 2004 Published by Elsevier B.V.

During reactor processing, fission products, among which iodine, are implanted by recoil inside the Zircaloy cladding tube. At the same time, oxidation of the cladding tube occurs, hence in the waste storage phase, zirconia acts as a migration barrier. Before chemical separation, the cladding tubes are sectioned into pieces called hulls in order to release the UO2 pellets which are rendered soluble in nitric acid. The hulls are collected as a solid waste and were embedded inside a concrete structure until 1995. In the perspective of geological storage, a great interest is given to iodine release processes in order to model and to extrapolate them to large time scales. In order to analyse the mechanisms involved in iodine migration, iodine atoms were introduced in Zircaloy oxidized samples by means of ion implantation. Corrosion tests were performed in order to simulate the impact of infiltrated water in the concrete disposal. Iodine release was measured by Rutherford backscattering spectroscopy. Processes responsible for iodine release are analysed.

Nuclear fuel devices of pressurised water reactors are composed of uranium oxide pellets which are enclosed in zircaloy-cylinders. During reactor operation in the contact with the fuel the zircaloy oxidises non-uniformly in depths of the order of some micrometers. Furthermore, energy deposition of fission recoils leads to sputtering of uranium onto the inner surface of the cladding material. Thus, sputtered uranium ions start to migrate outwards. Experiments were performed at Institut Laue Langevin (ILL) on the Lohengrin mass spectrometer in order to simulate the alteration of cladding tubes in contact with the nuclear fuel during irradiation. A thin UO2 deposition in direct contact with a zirconium foil was put into the ILL high flux reactor (5×1014 neutrons cm−2 s−1). The variation of energy loss of selected fission products in the foil put in evidence a zirconium oxidation whose kinetics constant is deduced from these experiments. Following oxidation the fission product mean kinetic energy is stabilised, but still a broadening of the energy distribution is observed, which is characteristic of uranium diffusion inside the zirconia target. Using Fick's model, a uranium diffusion coefficient in zirconia under irradiation is deduced.

Zirconia produced by the oxidation of zirconium alloys in nuclear reactors exhibits a phase transition under ionic irradiation, simulating a neutron irradiation. To understand the mechanism responsible for this irradiation driven phase transition, different kinds of projectiles were used to irradiate pure monoclinic zirconia samples. The evolution of these irradiated samples as a function of dpa has been studied using grazing X-ray diffraction. The Rietveld method has been applied on collected X-ray diffraction diagrams to study the phase produced under irradiation and the kinetics of its formation. Even at high dpa values, only the monoclinic and tetragonal phases were used to simulate X-ray diffraction diagrams. No amorphisation of zirconia was observed. The evolution of unit cells and short range strains in both phases under irradiation leads us to think that the irradiation driven transition is martensitic. Supposing that the inelastic stopping power in sub-cascades is responsible for the irradiation driven phase transition, we propose a model based on the Landau–Ginzburg effective hamiltonian to explain both the m→t transition observed under irradiation and the t→m transition measured during isochronal annealing after irradiation.

This paper is devoted to the study of apatites and zirconia used as conditioning materials of nuclear wastes. Among the long-lived fission products I and Tc release has to be evaluated. First, the stable isotopes Re and Mo considered as chemical homologous of Tc were implanted in hydroxyapatite. Then, cladding tube pieces (hulls) present a thin superficial zirconia layer in which most of the radioactivity is concentrated. Stable I release due to diffusion or to corrosion process was investigated in order to model its long term behaviour. Diffusion studies were performed by successive air annealings (300-900°C). The distribution evolution characterised by RBS allowed to identify the migration mechanisms. XPS and XANES methods were used to follow the chemical evolution of the compounds as function of temperature.

In pressurised water reactors the cladding tubes in zircaloy-4 are oxidised up to several micrometres on the internal face by direct contact with the UO2 pellets. At the same time, fission products such as 129I are implanted by recoil. Until 1995, the cladding tube pieces called hulls were embedded in concrete. The concrete medium being very basic, this study simulates the corrosion of hulls in the perspective of a long-term disposal in these severe pH conditions. This paper is dedicated to the study of the partial dissolution of the oxide layer, which is the responsible mechanism for the activity release. In order to follow the solid–liquid interface, europium was implanted into the material surface as a marker. Next, the corrosion induced in autoclave at 300°C, 140 bars and in alkaline water was studied. The determination of europium profiles using Rutherford backscattering spectrometry analysis (RBS) allows to deduce the fraction of dissolved ZrO2. It was shown that this dissolution is not homogeneous in porous zirconia and gives rise to the formation of crevices. Using the nuclear microprobe of Pierre Süe laboratory, a more precise study of the specimen surface was performed.

Nuclear waste ceramic forms among which the apatite, are under development as an alternative to waste glass in case of selective confinement. In that context, we studied the diffusion of lanthanide ions (La3+, Eu3+) in hydroxyapatite over a temperature range of storage interest, taking into account a possible enhanced diffusion due to irradiation effects. The lanthanide ions are introduced in apatite targets using ion implantation. The diffusion coefficients are deduced from Rutherford backscattering spectroscopy (RBS) at each step of annealing and irradiation procedure. Evidence of enhanced diffusion is shown and can be explained as a diffusion process governed by defect migration towards the surface. Time resolved laser-induced fluorescence measurements show that, during enhanced diffusion performed under vacuum, the europium ions substitute the calcium ions preferentially in Ca(I) hydroxyapatite sites.

Zirconia, ZrO2, produced by the oxidation of zirconium alloys in nuclear reactors, possesses a high stability under neutron irradiation. No amorphisation of yttrium-stabilised zirconia has been observed even at high dpa values (≈100 dpa). In pure monoclinic zirconia, a phase transition monoclinic → cubic (tetragonal) induced by irradiation has already been observed. The aim of this work is to study in detail the mechanism responsible for this transition. For that purpose, different kinds of irradiations with electrons (to study point defects) and low energetic ions (to study clusters due to collision cascades) have been performed on zirconia samples. A local probe (Raman spectroscopy) and a non-local probe (grazing X-ray diffraction) have been used to characterise the phase formed during irradiation, which is clearly the tetragonal phase. For the ionic implantation, the grazing X-ray diffraction permits to separate effects due to the ballistic collisions and the implantation peak. Using this method, it was possible to show that the profile of the tetragonal phase was only linked to the dpa profile. This result associated to the results obtained by the Raman spectroscopy (broadening of Raman peaks) shows that the phase transition may be induced by clusters formed near the collision cascades.

This study is related to the back end of the nuclear fuel cycle. During nuclear reactor operation, the inner side of the zircaloy cladding tube is implanted by recoil with fission products. Among them,I-129 poses a real problem firstly, because it is a volatile element and secondly because it has a very long half-life (T = 1.59 x 10(7) years). The aim of this paper is to make an analysis of iodine diffusion into zirconium compared to zirconium oxide. In order to analyse the mechanisms involved in iodine migration, stable (I-127) and radioactive iodine (I-131) species were implanted into zirconium. Diffusion profiles of iodine were followed as a function of successive annealing times (up to several hours) and temperatures (in the range 400-600 degrees C), either by using Rutherford backscattering spectrometry to profile I-127 Or by gamma spectroscopy to measure I-131 release. These two techniques allowed us to determine very low iodine diffusion coefficient values (down to 10(-17) cm(2) s(-1)) and to extrapolate them to waste storage conditions. (C) 2000 Elsevier Science S.A. All rights reserved.

During reader operation Zirconium cladding tubes enclosing the nuclear fuel are subject to various processes: penetration by fission fragments, oxidation due to contact with the Uranium-Oxide fuel and to the partial Oxygen pressure, and coating by Uranium and Higher Actinides bred in the fuel and deposited by the effect of sputtering onto the Zr surface. Oxidation of the Zirconium can be observed on-line by time dependent recording of fission fragment kinetic energy distributions at the mass separator Lohengrin at the ILL.

This study is related to the nuclear waste management and to the long-term behaviour of radionuclides in deep storage. The aim of this paper is to study the corrosion of zirconium in deep storage. In order to simulate the inner side of zircaloy cladding tubes, the zirconium samples were oxidised in air at 500°C during 3 h. Two hundred keV-implanted europium ions were used as marker to follow the evolution of the ZrO2–liquid interface. In a first step, air annealings were performed in the temperature range 850–950°C. Diffusion profiles were measured using 3 MeV alpha-particles Rutherford backscattering spectrometry (RBS) at each step of annealing. By extrapolation it is shown that at 300°C the europium diffusion in ZrO2 is negligible. In a second step, the specimens were exposed to a basic solution (pH 13.5) at 300°C in autoclave under a 107 Pa pressure. By means of RBS, it was possible to deduce the fraction of dissolved ZrO2. It was shown that this dissolution is inhomogeneous. These results are confirmed by scanning electron microscopy (SEM).