La reproduction expérimentale de transitoires thermiques accidentels de type Accident par Perte de Réfrigérant Primaire (APRP) en laboratoire a permis d’observer la fragmentation du combustible fortement irradié lorsque la gaine se déforme sous l’augmentation de la température. Ces fragments de petites tailles peuvent se relocaliser dans le ballon voire être éjectés hors du crayon cas de rupture de gaine. La zone High Burnup Structure (HBS) des combustibles fortement irradiés est la plus susceptible de se fragmenter et d’être relocalisée par sa position en périphérie de pastille. Pour expliquer ce phénomène, l’hypothèse retenue est que le transitoire provoque une surpression dans les bulles HBS ce qui mène à la décohésion des joints de grains et à la fragmentation. Cette thèse a pour but de développer un critère de fissuration mécanique du combustible pour mieux comprendre le comportement des bulles HBS lors des conditions thermiques APRP. Ce travail se base sur une méthode une méthode micromécanique en trois étapes : i) la représentation qui permet de caractériser la microstructure de la zone HBS (leurs dimensions, leur fraction volumique, et la pression interne). Deux sources d’informations seront utilisées : les observations expérimentales provenant de disques ou de pastilles de combustible irradiés à fort taux de combustion et d’outils numériques(avec Alcyone-Caracas [JSB+14]) / Under Loss Of Coolant Accident(LOCA) transients conditions, the high irradiated fuel is fragmented in small sizes fragments who can be relocated in the balloon, or being ejected out of the fuel rod if the latter burst. This work focuses on the pellet rim, where bubbles density increases owing to a higher irradiation level. Usually the hypothesis used to explain fuel fragmentation during transient is grain cleavage induced by over pressurized fission gas bubbles, located at the grain boundary. The aim of this study is to define a macroscopic fragmentation model based on a micro mechanical approach to have a better understanding of the fuel mechanical behaviour at lower scale : size and volume fraction of fragments. This PhD introduces a stepwise micromechanical method based on three steps : i) firstly, we detail how to model the HBS microstructure including pressurized porosities, based on experimental or numerical data and define a representative volume element (RVE)

Dans le cadre des études visant à garantir l'intégrité de la gaine du crayon combustible, EDF est amené à caractériser la ductilité de la gaine après un Accident de Perte de Réfrigérant Primaire (APRP). La thèse porte sur la caractérisation du comportement à rupture des gaines en Zircaloy-4 détendu pour lesquels les conditions d'APRP ont été simulées en laboratoire par une oxydation à haute température suivie d'un refroidissement. L'oxydation est effectuée à 1100°C et à 1200°C pour différentes durées ce qui conduit à des niveaux d'oxydation de 3% à 30% d'ECR (Equivalent Cladding Reacted). Deux types de refroidissement sont mis en oeuvre : la trempe à l'eau et le refroidissement à l'air. Les gaines oxydées comportent deux couches fragiles, la couche de zircone externe ZrO2 et la couche α(O), et une couche présentant une ductilité résiduelle, la couche ex-β.Les gaines oxydées ont fait l'objet de caractérisations en microscope optique, par analyse à la microsonde et par nano-indentation. Une corrélation entre la teneur en oxygène et la nano-dureté et le module d'Young a été proposée.L'essai Expansion due à la Compression (EDC) a été développé avec une instrumentation par stéréo-corrélation d'images puis a été utilisé pour caractériser le comportement mécanique des gaines oxydées. Le comportement des gaines oxydées a été étudié à partir de l'analyse des courbes macroscopiques de l'essai EDC et à partir des observations des échantillons rompus ou pré-déformés.Un scénario de rupture des gaines oxydées a été proposé. Ce scénario a été validé d'une part par la réalisation d'essais sur gaines sablées ne comportant que la couche ex-β et d'autre part par la modélisation de l'essai par la méthode des éléments finis. Un critère de rupture des gaines oxydées a par ailleurs été établi. La modélisation du comportement et le critère de rupture proposés ont été validés par la modélisation des essais de compression d'anneau. / In order to guarantee the integrity of nuclear fuel rod cladding, it is necessary for EDF to characterize the ductility of cladding after a Loss of Coolant Accident (LOCA). The thesis is about the characterization of the fracture behavior of cold-worked stress-relieved Zircaloy-4 claddings which have undergone LOCA conditions simulated in laboratory by a high temperature oxidation followed by a cooling. The high temperature oxidation is carried out at 1100°C and 1200°C with different times, which leads to different oxidation levels varying from 3% to 30% ECR (Equivalent Cladding Reacted). The high temperature oxidation is followed by two types of cooling: water quench and air cooling. The oxidized claddings contain two fragile layers - the outer zirconium oxide ZrO2 layer and the middle α(O) layer, and a layer which can have residual ductility - the inner ex-β layer.Characterizations by means of optical microscopy, electron probe micro analysis and nano-indentation have been carried out on the oxidized claddings. A correlation between the oxygen concentration and the nano-hardness and the Young's modulus has been proposed.The Expansion Due to Compression (EDC) test has been developed with an instrumentation of stereo digital image correlation, and then used to characterize the mechanical behavior of the oxidized claddings. The behavior of the oxidized claddings has been studied via macroscopic EDC test curves and observations of fractured or pre-deformed test samples.A fracture scenario of the oxidized claddings has been proposed. The fracture scenario has then been validated via EDC tests on oxidized claddings whose ZrO2 and α(O) layers have been removed, and via finite element modeling of EDC tests. Moreover, a fracture criterion has been established. The mechanical behavior modeling and the proposed fracture criterion have been validated by modeling of ring compression test.

In the event of hypothetical accident scenarios in PWR, emergency strategies have to be mapped out, in order to guarantee the reliable removal of decay heat from the reactor core, also in case of component breakdown. One essential passive heat removal mechanism is the reflux condensation cooling mode. This mode can appear for instance during a small break loss-of-coolant-accident (LOCA) or because of loss of residual heat removal (RHR) system during mid loop operation at plant outage after the reactor shutdown. In the scenario of a loss-of-coolant-accident (LOCA), which is caused by the leakage at any location in the primary circuit, it is considered that the reactor will be depressurized and vaporization will take place, thereby creating steam in the PWR primary side. Should this lead to ¿reflux condensation¿, which may be a favorable event progression, the generated steam will flow to the steam generator through the hot leg. This steam will condense in the steam generator and the condensate will flow back through the hot leg to the reactor, resulting in counter-current steam/water flow. In some scenarios, the success of core cooling depends on the behaviour of this counter-current flow. Over several decades, a number of experimental and theoretical studies of counter-current gas¿liquid two-phase flow have been carried out to understand the fundamental aspect of the flooding mechanism and to prove practical knowledge for the safety design of nuclear reactors. Starting from the pioneering paper of Wallis (1961), extensive CCFL data have been accumulated from experimental studies dealing with a diverse array of conditions A one-dimensional two field model was developed in order to predict the counter-current steam and liquid flow that results under certain conditions in the cold leg of a PWR when a SBLOCA (small break loss of coolant accident) in the hot leg is produced. The counter-current model that has been developed can predict the pressure, temperature, velocity profiles for both phases, also by taking into account the HPI injection system in the cold leg under a counter-current flow scenario in the cold leg. This computer code predicts this scenario by solving the mass, momentum and energy conservation equations for the liquid and for the steam separately, and linking them by using the interfacial and at the steam wall condensation and heat transfer, and the interfacial friction as the closure relations. The convective terms which appear in the discretization of the mass and energy conservation equations, were evaluated using the ULTIMATE-SOU (second order upwinding) method. For the momentum equation convective terms the ULTIMATE-QUICKEST method was used. The steam-water counter-current developed code has been validated using some experimental data extracted from some previously published articles about the direct condensation phenomenon for stratified two-phase flow and experimental data from the LAOKOON experimental facility at the Technical University of Munich. / Romero Hamers, A. (2014). STUDY OF THE THERMAL STRATIFICATION IN PWR REACTORS AND THE PTS (PRESSURIZED THERMAL SHOCK) PHENOMENON [Tesis doctoral]. Editorial Universitat Politècnica de València. https://doi.org/10.4995/Thesis/10251/36536 / Alfresco

The main target of the master´s thesis is reviewing the generation III nuclear reactors in term of the nuclear safety. At first we have to learn some theory of the nuclear safety in order to understand safety systems of the generation III nuclear reactors. Therefore the thesis is divided into two parts. Legislative and technical approaches to nuclear safety are mentioned in the first part. Regulatory bodies, whose task is to supervise nuclear safety in the nuclear power plants, belongs to the legislative approaches. There are defined terms such as defence in depth, redundancy, diversity, etc. There are mentioned methods to assessing nuclear safety – deterministic and probabilistic methods, especially probabilistic methods, for which a simple example is provided. There are also mentioned active and passive safety systems and their significance for nuclear safety and inherent safety too. There is an example of the function of the active and passive safety systems of the EDU nuclear power plant in conclusion of this issue. The second part deals with description of the selected nuclear reactors in context of the construction of the new units of nuclear power plant in Temelín. The nuclear reactors from companies, which applied for the public tender opened by ČEZ, a. s., for the construction of the ETE 3+4. Thus, the nuclear reactor MIR-1200 by ATOMSTROYEXPORT (Russian Federation), the nuclear reactor AP1000 by WESTINGHOUSE (USA) and the nuclear reactor EPR by AREVA (France) are taken into account . Comparison of the generation II and these generation III+ nuclear reactors necessarily belongs to this master´s thesis. These the generation III+ nuclear reactors are compared with the nuclear reactor VVER 440 (EDU) and in particular with the nuclear reactor VVER 1000, which is operated in the nuclear power plant Temelín. The final chapter contains generally appraisal of the whole problem.

The nuclear industry has long relied upon bounding parametric analyses in predicting the safety margins of reactor designs undergoing design-basis accidents. These methods have been known to return highly-conservative results, limiting the operating conditions of the reactor. The Best-Estimate Plus Uncertainty (BEPU) method using a modernized version of the Code-Scaling, Applicability, and Uncertainty (CSAU) methodology has been applied to more accurately predict the safety margins of the Oregon State University Advanced Plant Experiment (APEX) facility experiencing a Loss-of-Feedwater Accident (LOFA). The statistical advantages of the Bayesian paradigm of probability was utilized to incorporate prior knowledge when determining the analysis required to justify the safety margins. RELAP5 Mod 3.3 was used to accurately predict the thermal-hydraulics of a primary Feed-and-Bleed response to the accident using assumptions to accompany the lumped-parameter calculation approach. A novel coupling of thermal-hydraulic and statistical software was accomplished using the Symbolic Nuclear Analysis Package (SNAP). Uncertainty in Peak Cladding Temperature (PCT) was calculated at the 95/95 probability/confidence levels under a series of four separate sensitivity studies. / Graduation date: 2013

This work seeks to illustrate the potential benefits afforded by implementing aspects of fluid dynamics, especially the latest computational fluid dynamics (CFD) modeling approach, through numerical experimentation and the traditional discipline of physical experimentation to improve the calibration of the severe reactor accident analysis code, MELCOR, in one of several spent fuel pool (SFP) complete loss-of-coolant accident (LOCA) scenarios. While the scope of experimental work performed by Sandia National Laboratories (SNL) extends well beyond that which is reasonably addressed by our allotted resources and computational time in accordance with initial project allocations to complete the report, these simulated case trials produced a significant array of supplementary high-fidelity solutions and hydraulic flow-field data in support of SNL research objectives. Results contained herein show FLUENT CFD model representations of a 9x9 BWR fuel assembly in conditions corresponding to a complete loss-of-coolant accident scenario. In addition to the CFD model developments, a MATLAB based control-volume model was constructed to independently assess the 9x9 BWR fuel assembly under similar accident scenarios. The data produced from this work show that FLUENT CFD models are capable of resolving complex flow fields within a BWR fuel assembly in the realm of buoyancy-induced mass flow rates and that characteristic hydraulic parameters from such CFD simulations (or physical experiments) are reasonably employed in corresponding constitutive correlations for developing simplified numerical models of comparable solution accuracy. / text