Le projet ASTRID, réacteur nucléaire français de 4ème génération refroidi au sodium, est en cours de développement par le CEA. Dans ce projet, le développement de techniques de surveillance est identifié comme un problème majeur pour augmenter la sécurité du réacteur. L'utilisation de techniques de mesure par ultrasons est considérée comme un puissant outil d'inspection en raison de l'opacité du sodium liquide.A l'intérieur d'un circuit de refroidissement, l'hétérogénéité du milieu se produit du fait de l'état d'écoulement complexe, et les effets de cette hétérogénéité sur la propagation des ondes acoustiques ne sont pas négligeables. Ainsi, il est nécessaire d'effectuer des expériences de vérification, sachant que de telles expériences peuvent être des expériences à grande échelle. C'est pourquoi les méthodes de simulation numérique sont essentielles. L'objectif de l'étude de ma thèse est à appliquer la technique numérique des éléments spectraux, qui peut modéliser nos objets d'étude de manière plus précise que les méthodes de simulation plus classiques. Nous étudierons d'abord le potentiel de développement de la thermométrie ultrasonique similaire à celui d'un réacteur rapide refroidi au sodium avec simulation 2D. Un processus aléatoire Gaussien aura appliqué pour générer une fluctuation de la température.Afin d'étudier l'hétérogénéité en 3D et des champs de température plus réalistes dans le milieu, nous effectuerons une seconde étude numérique. Pour représenter l'hétérogénéité du sodium liquide, nous appliquerons un champ de température 4D (3D spatiale et 1D temporelle) calculé par modélisation numérique en dynamique des fluides avec LES réalisée par CEA STMF. / The ASTRID project, a french sodium-cooled nuclear reactor of 4th generation, is currently under development by the french alternative energies and atomic energy center (CEA). In this project, development of monitoring techniques is identified as an important issue to improve the plant safety. The use of ultrasonic measurement techniques is regarded as a powerful inspection tool due to the opacity of liquid sodium. Inside a cooling circuit, heterogeneity of the medium occurs because of a complex flow state, and then the effects of this heterogeneity on acoustic wave propagation are not negligible. Thus, it is necessary to carry out verification experiments, and such kind of experiments using liquid sodium may be large-scale. This is a reason why numerical simulation methods are essential. The objective of the study in the thesis is to apply a 3D spectral-element method, that we will show to be suitable to our targets more accurately than more classical numerical simulation methods.We will first study the development potential of ultrasonic thermometry in a liquid fluctuating sodium environment similar to that present in a sodium-cooled fast reactor with 2D simulation. Gaussian random process will be applied to generate fluctuations of temperature. To investigate 3D heterogeneity and more realistic temperature fields in the medium, in a second part of the thesis we will carry out a numerical study for 3D models of the reactor core. To represent the heterogeneity of liquid sodium, a four-dimensional temperature field (three spatial and one temporal dimension) calculated by computational fluid dynamics based on a large-eddy simulation performed by CEA STMF will be applied.

Thesis. 1976. Ph.D.--Massachusetts Institute of Technology. Dept. of Nuclear Engineering. / Microfiche copy available in Archives and Science. / Includes bibliographical references. / by Pascal De Laquil, III. / Ph.D.

This thesis is a follow up on "SPARC fast reactor design: Design of two passively metal-fuelled sodium-cooled pool-type small modular fast reactors with Autonomous Reactivity Control" by Tobias Lindström (2015). In this thesis the two reactors designed by Lindström in said thesis were evaluated. The goal was to determine the reactors ability to load follow as well as the burnup of the neutron absorber used in the passive control system. To be able to determine the dynamic behaviour of the reactors the reactivity feedbacks of the cores were modelled using Serpent, a Monte Carlo simulation software for 3D neutron transport calculations. These feedbacks were then implemented into a dynamic simulation of the core, primary and secondary circulation and steam generator. The secondary circulation and feedwater flow were used to regulate steam temperature and turbine power. The core was left at constant coolant flow and no control rods were used. The simulations showed that the reactor was able to load follow between 100 % and 40 % of rated power at a speed of 6 % per minute. It was also shown that the reactor could safely adjust its power between 100 % and 10 % of rated power suggesting that load following is possible below 40 % of rated power but at a lower speed. Finally the reactors were allowed compensate for the variations in a week of the Latvian wind power production in order to show one possible application of the reactor.

A series of accident simulations were performed using INL's thermal hydraulics code RELAP5-3D to analyze steady-state and transient behavior of a sodium cooled fast reactor. The reactor chosen for this study was General Electric's S-PRISM, which is a 1,000 MWt pool-type sodium-cooled fast reactor, designed for either an Oxide or Metal fueled core. Once key core characteristics including power profiles, reactivity feedback coefficients and delayed neutron parameters were calculated, S-PRISM was redesigned for a Nitride fueled core to take advantage of the Nitride fuel's high thermal conductivity and melting temperature. Loss of flow, loss of heat sink, loss of power and inadvertent control rod withdrawal accidents were simulated for each core at beginning, middle and end of cycle to determine if one fuel type provides significant safety advantages over the others.

Made available in DSpace on 2014-10-09T12:26:10Z (GMT). No. of bitstreams: 0 / Made available in DSpace on 2014-10-09T14:10:22Z (GMT). No. of bitstreams: 1 00951.pdf: 4937591 bytes, checksum: 160731d29ec9edf1fc78d0034f24638b (MD5) / Thesis (Doctorate) / IPEN/T / Massachusetts Institute of Technology - Cambridge, Mass - MIT

Made available in DSpace on 2014-10-09T12:26:10Z (GMT). No. of bitstreams: 0 / Made available in DSpace on 2014-10-09T14:10:22Z (GMT). No. of bitstreams: 1 00951.pdf: 4937591 bytes, checksum: 160731d29ec9edf1fc78d0034f24638b (MD5) / Thesis (Doctorate) / IPEN/T / Massachusetts Institute of Technology - Cambridge, Mass - MIT

The need for energy is growing in the world and the market of nuclear power is now once more expanding. Some issues of the current light-water reactors can be solved by the next generation of nuclear power, Generation IV, where sodium-cooled reactors are one of the candidates. Phénix was a French prototype sodium-cooled reactor, which is seen as a success. Although it did encounter an earlier unexperienced phenomenon, A.U.R.N., in which a negative reactivity transient followed by an oscillating behavior forced an automatic emergency shutdown of the reactor. This phenomenon lead to a lot of downtime of the reactor and is still unsolved. However, the most probable cause of the transients is radial movements of the core, referred to as core-flowering. This study has investigated the available documentation of the A.U.R.N. events. A simplified model of core-flowering was also created in order to simulate how radial expansion affects the reactivity of a sodium-cooled core. Serpent, which is a Monte-Carlo based simulation code, was chosen as calculation tool. Furthermore, a model of the Phénix core was successfully created and partly validated. The model of the core has a k_eff = 1.00298 and a neutron flux of (8.43+-0.02)!10^15 neutrons/cm^2 at normal state. The result obtained from the simulations shows that an expansion of the core radius decreases the reactivity. A linear approximation of the result gave the relation: change in k_eff/core extension = - 60 pcm/mm. This value corresponds remarkably well to the around - 60 pcm/mm that was obtained from the dedicated core-flowering experiments in Phénix made by the CEA. Core-flowering can recreate similar signals to those registered during the A.U.R.N. events, though the absence of trace of core movements in Phénix speaks against this. However, if core-flowering is the sought answer, it can be avoided by design. The equipment that registered the A.U.R.N. events have proved to be insensitive to noise. Though, the high amplitude of the transients and their rapidness have made some researcher believe that the events are a combination of interference in the equipment of Phénix and a mechanical phenomenon. Regardless, the origin of A.U.R.N. seems to be bound to some specific parameter of Phénix due to the fact that the transients only have occurred in this reactor. A safety analysis made by an expert committee, appointed by CEA, showed that the A.U.R.N. events are not a threat to the safety of Phénix. However, the origin of these negative transients has to be found before any construction of a commercial size sodium-cooled fast reactor can begin. Thus, further research is needed.

Les futurs réacteurs à neutrons rapides refroidis au sodium (RNR-Na) doivent remplir les critères GEN-IV à savoir présenter des qualités d'économie, de sûreté améliorée, de résistance à la prolifération et de minimisation des déchets. Ce projet de thèse est dédié à l'étude de l'impact des combustibles innovants (spécialement le combustible oxyde sphere-pac chargé en actinides mineurs) sur les performances de sûreté des RNR-Na dédiés à la transmutation.Le code de calcul SIMMER-III, développé à l'origine pour les phases avancées d'un accident grave, est utilisé pour les simulations. Ce code a été étendu dans le cadre de cette thèse afin d'améliorer la simulation de la phase primaire de l'accident, en introduisant le traitement des effets en réactivité liés à la dilatation du cœur et les spécificités du combustible sphere-pac (conductivité thermique, gap). Les transitoires complets (de la phase d'initiation aux phases avancées) sont simulés avec cette version étendue du code. Dans le cadre de cette thèse, les propriétés thermiques du combustible sphere-pac ont été modélisées et adaptées à SIMMER. Une méthodologie innovante tenant compte des effets en réactivité liés à la dilation thermique du cœur dans un maillage Eulérien et dans le cadre de la cinétique spatiale a ensuite été développée. A chaque pas de temps, les dimensions et densités dilatées sont calculées pour chaque cellule suite aux variations de températures. Des facteurs correctifs sont appliqués aux densités dilatées pour obtenir une configuration équivalente (en réactivité) ayant les dimensions non-dilatées et des densités modifiées. De nouvelles sections efficaces sont calculées à partir de ces densités et l'effet en réactivité lié à la dilatation est calculé. Les résultats sont prometteurs pour des dilatations uniformes et non-uniformes. Des limitations dans le cas de dilatations non-uniformes ont été identifiées et des calculs neutroniques ont été effectués en vue de futurs développements SIMMER. Les résultats préliminaires sont encourageants. Enfin, deux cœurs RNR-Na, issus du précédent projet CP-ESFR, ont été modélisés avec des combustibles sphere-pac : le Working Horse et le CONF2 (présentant un plénum sodium élargi pour une diminution de l'effet de vide sodium). Des analyses de sûreté ont été effectuées afin de fournir une première évaluation du comportement du combustible sphere-pac comparé au combustible pastille. Les deux options sont analysées en situation nominale et accidentelle (accident de perte de débit primaire) en début de vie du cœur et après irradiation. Les analyses révèlent deux phases à considérer en début de vie pour le combustible sphere-pac. Au démarrage du réacteur, ce combustible n'est pas restructuré et sa conductivité thermique est très inférieure à celle du combustible pastille. Après quelques heures sous irradiation, il se restructure suite aux importants gradients de température, ce qui améliore sa conductivité. Il se comporte alors de façon similaire au combustible pastille. Ce travail a également permis d'évaluer le comportement accidentel du cœur CONF2 qui subit un transitoire doux, prouvant que le large plénum sodium prévient efficacement de larges insertions de réactivité positive. Cependant, avec l'ajout d'américium ou suite à l'irradiation, des excursions de puissance et de réactivité plus prononcées sont observées. Ce travail a permis de démontrer que le combustible sphere-pac ne semble pas causer de problèmes de sûreté spécifiques comparé au combustible pastille, dans les conditions de simulations actuelles. La prise en compte des effets en réactivité liés à la dilatation du cœur avec cette version étendue de SIMMER retarde et réduit le potentiel énergétique lors d'un accident. Les analyses confirment également l'action atténuante du plénum sodium sur les transitoires conduisant à la vidange du sodium du coeur. Le comportement du combustible sphere-pac dans ces conditions ouvre une perspective à son utilisation en RNR-Na. / Future sodium cooled fast reactors (SFRs) have to fulfill the GEN-IV requirements of enhanced safety, minimal waste production, increased proliferation resistance and high economical potential. This PhD project is dedicated to the evaluation of the impact of innovative fuels (especially minor actinides bearing oxide sphere-pac fuels) on the safety performance of advanced SFRs with transmutation option. The SIMMER-III code, originally tailored to mechanistically analyze later phases of core disruptive accidents, is employed for accident simulations. During the PhD project, the code has been extended for a better simulation of the early accident phase introducing the treatment of thermal expansion reactivity effects and for taking into account the specifics of sphere-pac fuels (thermal conductivity and gap conditions). The entire transients (from the initiating event to later accident phases) have been modeled with this extended SIMMER version. Within this PhD work, first the thermo-physical properties of sphere-pac fuel have been modeled and casted into SIMMER-III. Then, a new computational method to account for thermal expansion feedbacks has been developed to improve the initiation phase modeling of the code. The technique has the potential to evaluate these reactivity feedbacks for a fixed Eulerian mesh and in a spatial kinetics framework. At each time step, cell-wise expanded dimensions and densities are calculated based on temperature variations. Density factors are applied to the expanded densities to get an equivalent configuration (in reactivity) with original dimensions and modified densities. New cross sections are calculated with these densities and the reactivity of the equivalent configuration is computed. The developed methods show promising results for uniform and non-uniform expansions. For non-uniform expansions, model improvement needs have been identified and neutronics simulations have been carried out to support future SIMMER extensions. Preliminary results are encouraging. In the third part of the PhD, two core designs with conventional and sphere pac fuels are compared with respect to their transient behavior. These designs were established in the former CP-ESFR project: the working horse core and the optimized CONF2 core (with a large sodium plenum above the core for coolant void worth reduction). The two fuel design options are compared for steady state and transient conditions (Unprotected Loss of Flow accident, ULOF) either at beginning of life (BOL) or under irradiated conditions. Analyses for sphere-pac fuel reveal two main phases to consider at BOL. At start-up, the non-restructured sphere-pac fuel shows a low thermal conductivity compared to pellet fuel of same density. However, the fuel restructures quickly (in a few hours) due to the high thermal gradients and its thermal conductivity recovers. The fuel then shows a behavior close to the pellet one. The study also shows that the CONF2 core leads to a very mild transient for a ULOF accident at BOL. The large upper sodium plenum seems to effectively prevent large positive reactivity insertions. However, stronger reactivity and power peaks are observed under irradiated conditions or when americium is loaded in the core and lower axial blanket. This PhD work demonstrates, under current simulation conditions, that sphere-pac fuels do not seem to cause specific safety issues compared to standard pellet fuels, when loaded in SFRs. The accurate simulation of core thermal expansion reactivity feedbacks by means of the extended SIMMER version plays an important role in the accident timing (simulations confirm the expected delay in the first power peak) and on the energetic potential compared to the case where these feedbacks are omitted. The analyses also confirm the mitigating impact of a large sodium plenum on transients with voiding potential. The behavior of sphere-pac fuel in these conditions opens a perspective to its practical application in SFRs.

Made available in DSpace on 2014-10-09T12:44:01Z (GMT). No. of bitstreams: 0 / Made available in DSpace on 2014-10-09T14:07:25Z (GMT). No. of bitstreams: 1 06864.pdf: 11106654 bytes, checksum: 851c7803db872d59fc1f49dc465fa8af (MD5) / Tese (Doutoramento) / IPEN/T / Instituto de Pesquisas Energeticas e Nucleares - IPEN/CNEN-SP

Made available in DSpace on 2014-10-09T12:44:01Z (GMT). No. of bitstreams: 0 / Made available in DSpace on 2014-10-09T14:07:25Z (GMT). No. of bitstreams: 1 06864.pdf: 11106654 bytes, checksum: 851c7803db872d59fc1f49dc465fa8af (MD5) / Tese (Doutoramento) / IPEN/T / Instituto de Pesquisas Energeticas e Nucleares - IPEN/CNEN-SP