Les travaux menés au cours de cette thèse s’inscrivent dans le cadre général de l’évaluation du terme source (relâchement de gaz de fission) en situation incidentelle de type APRP (Accident de Perte de Réfrigérant Primaire). Lors de tels transitoires thermiques, le relâchement de gaz de fission se fait par bouffées successives : une première entre 600°C et 800°C et la seconde à environ 1100°C. Ces travaux de thèse s’intéressent à cette première. Il semblerait que la bouffée à 600-800°C proviendrait du centre de la pastille combustible. L’objectif de cette thèse est d’étudier les mécanismes à l’origine de cette bouffée.Afin de mieux comprendre ces mécanismes, une étude a été menée sur l’ensemble des traitements thermiques réalisés dans la plateforme expérimentale MERARG. L’analyse de cette base de données a révélé 2 points importants : 1) Dans les conditions expérimentales de MERARG, aucune fagmentation significative du combustible n’est observée à des températures inférieures à 1000°C. 2) Le niveau de relâchement de gaz de fission entre 600°C et 800°C semble augmenter avec le temps d’entreposage du combustible.Le premier point indique que la fragmentation du combustible n’est pas une condition nécessaire au relâchement de gaz de fission dans cette gamme de température : d’autres mécanismes peuvent être à l’origine de ce relâchement. Durant l’entreposage, le combustible est soumis principalement à l’auto-irradiation α. Celle-ci a pour effet de créer des défauts dans une zone qui n’en contenait initialement pas. Nous avons démontré que la cinétique du relâchement de gaz de fission entre 600°C et 800°C est concomitante avec la cinétique de recuit de défauts d’autoirradiation α. De plus, une cinétique auto-catalytique de germination-croissance de nano-clusters de gaz a été développée et confrontée aux résultats expérimentaux. En outre, une étude sur matériaux simulants démontre que, sur des pastilles d’UO2 frittées et implantées en xénon, une irradiation en régime électronique a pour effet d’accroitre le relâchement entre 600°C et 800°C. La littérature décrit la remise en solution des bulles de gaz de fission sous l’effet d’une irradiation de ce type. De plus, lors de leur remise en solution, les gaz de fission s’insèrent dans les défauts de la structure cristalline. Lors d’un traitement thermique, le recuit des défauts entraine la mobilité des atomes de gaz de fission insérés de ces mêmes défauts. Par germination-croissance, les paires gaz/défauts rejoignent un chemin de sortie, les gaz de fission sont donc relâchés.Ce travail a donc permis de retenir l’hypothèse d’un mécanisme de relâchement de gaz de fission entre 600°C et 800°C par recuit de défauts sans fragmentation significative du combustible. / The subject of this thesis concerns the evaluation of the source term (fission gas release) in incidental situations of type LOCA (Loss of Coolant Accident) of nuclear fuels. During such thermal transients, the fission gas release is characterized by successive bursts : the first one occurring between 600 and 800°C, and a second one at about 1100°C. This work is about the first burst release. It appear that this one come from the centre of the fuel pellet. The aim of this thesis is to study the mechanisms responsible for the fission gas release between 600°C and 800°C.To this purpose, we collected more than 200 results of thermal treatments carried out using the experimental platform MERARG. The analysis of this database reveals two important results : under the experimental conditions of MERARG, no significant fragmentation of the fuel was observed at temperatures below 1000°C ; the amount of fission gas release between 600°C and 800°C appears to increase with fuel storage time.The first result suggests the fragmentation of the fuel is not a necessary condition for the release of fission gas in this temperature range. Other mechanisms may then be responsible for this gas release. During its storage, the fuel undergoes α particle self-irradiation. We demonstrate that the kinetics of fission gas release between 600°C and 800°C is simultaneous with the kinetics of the annealing of self-irradiation defects at this same temperature. The mechanism involves an autocatalytic process leading to a kinetic of fast germination-growth of gas nano-clusters. This model perfectly explains the experimental results in the database. To confirm this mechanism, a study on surrogate materials demonstrates that, in UO2 pellets sintered and implanted by Xe, irradiations in the electronic regime actually promote the release of implanted Xe at those temperatures. The re-dissolution of the fission gas bubbles by this kind of irradiation is consistent with observations in other contexts. Those conclusions allow to extend the mechanism for release to irradiated fuel.During the storage of the fuel, α self-irradiation promotes the re-dissolution of the trapped gas. The consequences of this effect are particularly important in the region close to the grain boundaries, where the concentration of defects is also larger. The irradiation mechanism increases the fraction of fission gas available for release, depleting the amount of gas initially trapped in bubbles. The gas in solution can effectively be carried by crystal defects, largely available in the irradiated fuel and whose migration at 600-800°C induces the mobility of the fission gas. When they reach an outlet path, the gas can be released from the pellet in a way consistent with the model of autocatalytic germination-growth we developed to explain the macroscopic results of the database.In conclusion, this work supports the hypothesis of a mechanism of fission gas release in the range 600-800°C via a mechanism involving the migration and annealing of defects without significant fragmentation of the fuel.

<p>Presently there is a clear trend of increasing demands on in-pile performance of nuclear fuel. Higher target burnups, part length rods and various fuel additives are some examples of this trend. Together with an increasing demand from the public for even safer nuclear power utilisation, this implies an increased focus on various experimental, preferably non-destructive, methods to characterise the fuel.</p><p>This thesis focuses on the development and experimental evaluation of such methods. In its first part, the thesis presents a method based on gamma-ray spectroscopy with germanium detectors that have been used at various power reactors in Europe. The aim with these measurements is to provide information about the thermal power distribution within fuel assemblies in order to validate core physics production codes. The early closure of the Barsebäck 1 BWR offered a unique opportunity to perform such validations before complete depletion of burnable absorbers in Gd-rods had taken place. To facilitate the measurements, a completely submersible measuring system, LOKET, was developed allowing for convenient in-pool measurements to be performed.</p><p>In its second part, the thesis describes methods that utilise in-pile measurements. These methods have been used in the Halden test-reactor for determination of fission gas release, pellet-cladding interaction studies and fuel development studies.</p><p>Apart from the power measurements, the LOKET device has been used for fission gas release (FGR) measurements on single fuel rods. The significant reduction in fission gas release in the modern fuel designs, in comparison with older designs, has been demonstrated in a series of experiments. A FGR database covering a wide range of burnup, power histories and fuel designs has been compiled and used for fuel performance analysis. The fission gas release has been measured on fuel rods with average burnups well above 60 MWd/kgU. The comparison between core physics calculations (PHOENIX-4/POLCA-7) and the in-pool measurements of thermal power indicates that the nodal power can generally be predicted with an accuracy within 4% and the bundle power with an accuracy better than 2%, expressed as rms errors.</p><p>In-pile experiments have successfully simulated the conditions that occur in a fuel rod following a primary debris failure, being secondary fuel degradation. It was concluded that massive hydrogen pick-up takes place during the first few days following the primary failure and that a pre-oxidized layer does not function as a barrier towards hydriding in an environment with a very high partial pressure of hydrogen. Another series of in-pile experiments clearly indicate that increased UO₂ grain size is an effective way of suppressing fission gas release in LWR fuel up to the burnup level covered (55 MWd/kgUO₂).</p>

Presently there is a clear trend of increasing demands on in-pile performance of nuclear fuel. Higher target burnups, part length rods and various fuel additives are some examples of this trend. Together with an increasing demand from the public for even safer nuclear power utilisation, this implies an increased focus on various experimental, preferably non-destructive, methods to characterise the fuel. This thesis focuses on the development and experimental evaluation of such methods. In its first part, the thesis presents a method based on gamma-ray spectroscopy with germanium detectors that have been used at various power reactors in Europe. The aim with these measurements is to provide information about the thermal power distribution within fuel assemblies in order to validate core physics production codes. The early closure of the Barsebäck 1 BWR offered a unique opportunity to perform such validations before complete depletion of burnable absorbers in Gd-rods had taken place. To facilitate the measurements, a completely submersible measuring system, LOKET, was developed allowing for convenient in-pool measurements to be performed. In its second part, the thesis describes methods that utilise in-pile measurements. These methods have been used in the Halden test-reactor for determination of fission gas release, pellet-cladding interaction studies and fuel development studies. Apart from the power measurements, the LOKET device has been used for fission gas release (FGR) measurements on single fuel rods. The significant reduction in fission gas release in the modern fuel designs, in comparison with older designs, has been demonstrated in a series of experiments. A FGR database covering a wide range of burnup, power histories and fuel designs has been compiled and used for fuel performance analysis. The fission gas release has been measured on fuel rods with average burnups well above 60 MWd/kgU. The comparison between core physics calculations (PHOENIX-4/POLCA-7) and the in-pool measurements of thermal power indicates that the nodal power can generally be predicted with an accuracy within 4% and the bundle power with an accuracy better than 2%, expressed as rms errors. In-pile experiments have successfully simulated the conditions that occur in a fuel rod following a primary debris failure, being secondary fuel degradation. It was concluded that massive hydrogen pick-up takes place during the first few days following the primary failure and that a pre-oxidized layer does not function as a barrier towards hydriding in an environment with a very high partial pressure of hydrogen. Another series of in-pile experiments clearly indicate that increased UO2 grain size is an effective way of suppressing fission gas release in LWR fuel up to the burnup level covered (55 MWd/kgUO2).

The aim of this master thesis is to show a connection among nuclear fuel safety, experiments led in research reactors and calculation codes. This thesis focuses on the calculation code Transuranus. There are represented four experiments, which were calculated in Transuranus. The fission gas release, elongation and growth of fuel were particularly monitored. Is is possible to set differences among versions v1m1j09 and v1m3j12 from achieved results, as well as the influence of selected Transuranus parameters on the results, so the thesis may bring new pieces of knowledge for improvement of safety analysis calculation by Transuranus.