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Academic literature on the topic 'Corium'

Author: Grafiati

Published: 4 June 2021

Last updated: 8 June 2024

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Journal articles on the topic "Corium"

Irfan, Muhamad, Ismail Humolungo, Asril Pramutadi Andi Mustari, and Sidik Permana. "Comparison of Melted Corium Relocation during Severe Accident of High Temperature Reactor using Moving Particle Semi-Implicit Method." Computational And Experimental Research In Materials And Renewable Energy 6, no. 1 (May 31, 2023): 1. http://dx.doi.org/10.19184/cerimre.v6i1.39363.

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System failure in nuclear reactors can cause degradation of a reactor core, allowing melting and relocation of the corium to the lower plenum in the nuclear reactor system. In this study, a severe accident simulation was carried out using the Moving Particle Semi-Implicit (MPS) method. In this method, we model the relocation of molten corium on the reactor core (support plate) to the lower plenum for several conditions with variations: corium material, lower plenum conditions, temperature, viscosity, and density. Those treatments were carried out in order to be able to compare and analyze the characteristics of the corium melt by reviewing the velocity profiles. The formation of a corium pool and debris bed can result in significant temperature differences and high heat flux against the walls of the reactor vessel, causing a decrease in the integrity of the reactor vessel and reactor failure.Keywords: Corium, Uranium Dioxide (UO2), Zirconium Dioxide (ZrO2), Fluid Relocation, Moving Particle Semi-Implicit (MPS).

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Skakov, M. K., N. Ye Mukhamedov, I. I. Deryavko, and I. M. Kukushkin. "Thermal Properties and Phase Composition of Full-Scale Corium of Fast Energy Reactor." Key Engineering Materials 736 (June 2017): 58–62. http://dx.doi.org/10.4028/www.scientific.net/kem.736.58.

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This paper has studied the phase composition and determined thermal properties of full-scale fast power corium at a room temperature. The obtained data of the corium thermal properties can be used for calculating temperature fields during modeling the processes for retention of corium melting in the nuclear reactor core.

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Journeau, Christophe, Laurence Aufore, Léonie Berge, Claude Brayer, Nathalie Cassiaut-Louis, Nicolas Estre, Frédéric Payot, et al. "Corium-Sodium and Corium-Water Fuel-Coolant-Interaction Experimental Programs for the PLINIUS2 Prototypic Corium Platform." Nuclear Technology 205, no. 1-2 (July 18, 2018): 239–47. http://dx.doi.org/10.1080/00295450.2018.1479580.

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Skakov, M. K., V. V. Baklanov, K. O. Toleubekov, A. S. Akaev, M. K. Bekmuldin, and A. V. Gradoboev. "MODELING OF THE CORIUM AND METALS – COOLERS INTERACTION IN A CORE CATCHER OF A LIGHT WATER REACTOR." NNC RK Bulletin, no. 2 (July 6, 2023): 49–57. http://dx.doi.org/10.52676/1729-7885-2023-2-49-57.

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The core catcher is one of the mandatory elements of the reactor safety system, which prevents the release of reactor core materials in a severe accident. The core catcher is steel vessel filled with sacrificial materials (SM) and forming a tank where a corium melt coming from the core is formed. The trap is a steel body filled with sacrificial materials (LM) and forming a vessel where a corium bath is formed coming from the core. The melt formed in the core catcher is cooled by heat removal to the cooling water through the shell of the steel vessel, as well as by water supplied directly to the surface of the melt after the dissolution process of the SM in corium (gravitational inversion). The delay in the water supply to the melt is associated with the features of the component structure of corium and its interaction with water (the formation of explosive hydrogen and the possibility of its detonation, as well as the threat of a steam explosion). However, a certain amount of time is spent on the implementation of gravitational inversion, and it is desirable to start the water supply to the melt immediately at the moment when the corium enters the core catcher due to the danger of the system going beyond the permissible limits (the beginning of boiling of uranium dioxide) due to decay heat in the corium. In this regard, the authors have an idea – to use a fusible metal for additional cooling of the surface of the corium in order to organize heat removal and reduce the temperature of the corium in the period before the end of the gravitational inversion process. The article presents the results of modeling the interaction of corium with candidate low-melting metals – coolers. The modeling was conducted using the ANSYS software package. As a result of the conducted work, the time for which each of the considered cooling metals will reach the points of phase transitions of melting and boiling is determined. The analysis of the results allowed us to draw appropriate conclusions about the possible practical implementation of the proposed method of cooling corium.

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Skakov, М. K. "THE METHOD OF CORIUM COOLING IN A CORE CATCHER OF A LIGHT-WATER NUCLEAR REACTOR." Eurasian Physical Technical Journal 19, no. 3 (41) (September 22, 2022): 69–77. http://dx.doi.org/10.31489/2022no3/69-77.

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During the development of a severe accident at nuclearpower plantwith a core melting, corium is formed. One of the main barriers preventing outflow of corium into the environment is a melt localization device or a melt trap. The melt trap must accept and prevent the corium parameters from exceeding critical values, ensuring its retention in a controlled volume and cooling. For this reason, melt traps are subject to serious requirements regarding cooling methods to ensure effective containment of the melt in the core of a nuclear reactor. In the presented article, experimental studies of the interaction between corium and water, which was supplied to the surface of the corium in a melt trap for its cooling, were analyzed. As a result of the work, a number of significant problems associated with the low efficiency of this cooling method were identified, and possible ways to eliminate them were considered. A solution is proposed for optimizing the method of corium cooling in a melt trap, as well as for the scope of research on the possibility of implementing the proposed method in practice and analyzing its effectiveness using the VCG-135 test-bench and the Lava-B facility.

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Skakov, Mazhyn, Viktor Baklanov, Assan Akaev, Ivan Kukushkin, Maxat Bekmuldin, Kuanyshbek Toleubekov, Alexandr Gradoboev, and Olga Stepanova. "On the Possibility of Forming a Corium Pool by Induction Heating in a Melt Trap of the Lava-B Facility." Applied Sciences 13, no. 4 (February 15, 2023): 2480. http://dx.doi.org/10.3390/app13042480.

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This paper presents the results of computational and physical studies on the production of corium and its retention in an MR’s melt trap of the Lava-B facility. A feature of the Lava-B facility used in the IAE NNC RK to study the processes occurring during a severe accident at a nuclear reactor, is the separation of the stages of the reactor core corium formation and its interaction with structural materials. The melting of materials takes place in an induction furnace with a hot crucible, after which it moves to a melt receiver (MR) in which the test object is located. In the case of studies of processes occurring outside the reactor vessel, this is a special trap, which is placed in the inductor to simulate decay heat. However, based on the conservative computational estimates, it was found that the inductor power in the MR can be sufficient to directly produce, melt, and, subsequently, maintain the corium in the liquid phase. In this regard, in order to optimize the experiments under controlled conditions, the authors came up with the idea to experimentally test the possibility of producing corium by induction heating directly in the MR’s melt trap. In addition, according to the authors, this method would obviate the problem of corium contact with the carbon environment of the melting furnace of the Lava-B facility. Previously, burden heating simulating corium was modeled on the computer using available parameters of the MR’s induction heater. Based on the numerical experiment, the conditions for physical modeling of the corium production in the MR’s melt trap were established. An analysis of the physical modeling showed that during the burden heating in the melt trap, its metal components became liquid, thus, forming a melt pool. However, in terms of this design of the trap, there were problems associated with the complete melting of all corium components, as well as with the integrity of the experimental device when forming the corium pool and during the actual physical modeling.

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Skakov, Mazhyn, Viktor Baklanov, Maxat Bekmuldin, Ivan Kukushkin, Assan Akaev, Alexander Gradoboev, and Olga Stepanova. "Results of experimental simulation of interaction between corium of a nuclear reactor and sacrificial material (Al2O3) with a lead layer." AIMS Materials Science 11, no. 1 (2024): 81–93. http://dx.doi.org/10.3934/matersci.2024004.

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<abstract> This paper presents the results of an experimental study of the interaction of a candidate sacrificial material (SM) for a light water reactor melt trap with corium at the Lava-B test-bench. The candidate sacrificial material is a combination of aluminum oxide and a lead layer. The idea of using such a combination of SM is based on the fact that when the lead layer interacts with corium, there will be an increase in the intensity of heat removal from the corium, as well as the chemical interaction between the corium and SM due to the high heat-conducting properties of lead. This approach will improve the efficiency of corium localization in the melt trap compared to the current set of sacrificial material. Experiments have shown active melting and boiling of lead during its interaction with corium. This is confirmed both by the readings of thermocouples and by the X-ray diffraction phase analysis of the deposit material formed on the walls of the melt receiver (MR) of the Lava-B bench, sampled after the experiment. The experiment results show that the lead layer reduces the rate of increase in the temperature of the corium and increases the rate of erosion of the ceramic part of the SM. With these circumstances, it is possible to conclude that the use of aluminum oxide with a lead layer is promising in practice. </abstract>

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Skakov, M. K. "ANALYSIS OF METHODS FOR SIMULATING THE DECAY HEAT IN CORIUM WHEN MODELING A SEVERE ACCIDENTS AT NUCLEAR POWER PLANT." Eurasian Physical Technical Journal 21, no. 1 (47) (March 29, 2024): 57–66. http://dx.doi.org/10.31489/2024no1/57-66.

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It is known that during development of a severe accident at a nuclear power plant, the melting of core materials and theformation of corium occurs. A feature of corium is the presence of a decay heat, which contributes a lot to the nature of its interaction with the structural materials of the reactor facility. In this regard, quite serious requirements are imposed on methods for simulating decay heat in the corium prototype, which relate to both the uniformity of the volume distribution and its intensity. This paper presents a comparative analysis of existing methods for decay heat simulation in corium, which are used at various experimental facilities investigating the operation of passive protection systems in severe accidents with reactor meltdown at nuclear power plants. By comparing the advantages and disadvantages, a more practical method of decay heat simulation is determined and ways are proposed to further improve the chosen method to fully simulate the thermal field of a real corium.

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Becker, Joern-Martin, Doris Bulach, and Ulrich Müller. "Skora, corium, ledder." Hansische Geschichtsblätter 122 (January 13, 2021): 87–116. http://dx.doi.org/10.21248/hgbll.2004.166.

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Spitalny, Hans-Henning. "Corium transplantation cannula." Aesthetic Plastic Surgery 17, no. 2 (June 1993): 157–61. http://dx.doi.org/10.1007/bf02274737.

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Dissertations / Theses on the topic "Corium"

Quaini, Andrea. "Étude thermodynamique du corium en cuve - Application à l'interaction corium/béton." Thesis, Université Grenoble Alpes (ComUE), 2015. http://www.theses.fr/2015GREAI061/document.

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Lors d’un accident grave dans un réacteur nucléaire à eau pressurisée, le combustible nucléaire va réagir avec le gaines en Zircaloy, les absorbants neutroniques et les structures métalliques environnantes pour former un mélange partiellement ou complètement fondu. Ce cœur fondu peut ensuite interagir avec la cuve en acier du réacteur pour former un mélange appelé corium en cuve. Par la suite, le corium peut percer la cuve et venir se déverser sur le radier en béton en-dessous du réacteur. En fonction du scénario considéré, le corium qui va réagir avec le béton peut être constitué soit d’une seule phase liquide oxyde ou de deux liquides, métallique et oxyde. L’objectif de la thèse est l’étude de la thermodynamique du corium en cuve, prototypique U-Pu-Zr-Fe-O. L’approche utilisée est basée sur la méthode CALPHAD, qui permet de développer un modèle thermodynamique sur ce système complexe à partir de données expérimentales thermodynamiques et de diagramme de phases. Des traitements thermiques sur le système O-U-Zr ont permis de mesurer deux conodes dans la lacune de miscibilité à l’état liquide à 2567 K. De plus, des températures de liquidus ont été mesurées sur trois échantillons riches en Zr, en utilisant le montage de chauffage laser de l’ITU. Par la même méthode, des températures de solidus ont été obtenues sur le système UO2-PuO2-ZrO2. L’influence de l’atmosphère réductrice ou oxydante sur le comportement à la fusion de ce système a été étudiée pour la première fois. Les résultats montrent que la stœchiométrie en oxygène de ces oxydes dépend fortement du potentiel d’oxygène et de la composition en métal des échantillons. La lacune de miscibilité à l’état liquide a également été mise en évidence dans un échantillon U-O-Zr-Fe. L’ensemble de ces nouvelles données expérimentales avec celles de la littérature a permis de développer le modèle sur le système U-Pu-Zr-Fe-O. Pour tous les échantillons, des calculs de chemin de solidification avec ce modèle ont servi à interpréter les microstructures de solidification observées. Un bon accord est obtenu entre les calculs et les résultats expérimentaux. Des traitements thermiques sur deux échantillons de corium hors cuve ont permis de montrer l’influence de la composition du béton sur la nature des phases liquides formées à haute température. Les microstructures de solidification ont été interprétées à l’aide de calculs avec la base de données TAF-ID. En parallèle, un nouveau montage expérimental appelé ATTILHA, utilisant la lévitation aérodynamique et le chauffage laser, a été conçu et développé pour mesurer des données de diagramme de phase à haute température. Ce montage a été validé avec des systèmes oxydes bien connus. De plus, cette méthode a permis d’observer in-situ à l’aide de la caméra infra-rouge la formation de la lacune de miscibilité à l’état liquide dans le système O-Fe-Zr lors de l’oxydation d’une bille d’alliage Fe-Zr. La prochaine étape du développement est la nucléarisation du montage pour effectuer des mesures sur des échantillons contenant de l’uranium. La mise en place d’une caméra ultra rapide (5000 Hz) pour l’étude de propriétés thermo-physiques de mélanges de corium en cuve et hors cuve est également envisagée. La synergie entre le développement de ces outils expérimentaux et de calcul devrait permettre d’améliorer la description thermodynamique du corium et des codes de calcul sur les accidents graves utilisant ces données thermodynamiques
During a severe accident in a pressurised water reactor, the nuclear fuel can interact with the Zircaloy cladding, the neutronic absorber and the surrounding metallic structure forming a partially or completely molten mixture. The molten core can then interact with the reactor steel vessel forming a mixture called in-vessel corium. In the worst case, this mixture can pierce the vessel and pour onto the concrete underneath the reactor, leading the formation of the ex-vessel corium. Furthermore, depending on the considered scenario, the corium can be formed by a liquid phase or by two liquids, one metallic the other oxide. The objective of this thesis is the investigation of the thermodynamics of the prototypic in-vessel corium U-Pu-Zr-Fe-O. The approach used during the thesis is based on the CALPHAD method, which allows to obtain a thermodynamic model for this complex system starting from phase diagram and thermodynamic data. Heat treatments performed on the O-U-Zr system allowed to measure two tie-lines in the miscibility gap in the liquid phase at 2567 K. Furthermore, the liquidus temperatures of three Zr-enriched samples have been obtained by laser heating in collaboration with ITU. With the same laser heating technique, solidus temperatures have been obtained on the UO2-PuO2-ZrO2 system. The influence of the reducing or oxidising on the melting behaviour of this system has been studied for the first time. The results show that the oxygen stoichiometry of these oxides strongly depends on the oxygen potential and on the metal composition of the samples. The miscibility gap in the liquid phase of the U-Zr-Fe-O system has been also observed. The whole set of experimental results with the literature data allowed to develop the thermodynamic model of the U-Pu-Zr-Fe-O system. Solidification path calculations have been performed for all the investigated samples to interpret the microstructures of the solidified samples. A good accordance has been obtained between calculation and experimental results. Heat treatments on two ex-vessel corium samples showed the influence of the concrete composition on the nature of the liquid phases formed at high temperature. The observed microstructures have been interpreted by means of calculation performed with the TAF-ID database. In parallel, a novel experimental setup named ATTILHA based on aerodynamic levitation and laser heating has been conceived and developed to obtain high temperature phase diagram data. This setup has been validated on well-known oxide systems. Furthermore, this technique allowed to observe in-situ, by using an infrared camera, the formation of a miscibility gap in the liquid phase of the O-Fe-Zr system by oxidation of a Fe-Zr sample. The next step of the development will be the nuclearization of the apparatus to investigate U-containing samples. The implementation of a very fast visible camera (5000 Hz) to investigate the thermo-physical properties of in-vessel and ex-vessel corium mixtures is also underway. The synergy between the development of experimental and calculation tools will allow to improve the thermodynamic description of the corium and the severe accident code using thermodynamic input data

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Zabiégo, Magali. "Rayonnement d'un bain de corium dans un milieu chargé en aérosols issus de l'interaction corium/béton." Aix-Marseille 1, 1996. http://www.theses.fr/1996AIX11002.

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Le cas hypothetique de la perte de refrigerant primaire dans un reacteur a eau pressurisee (rep) peut entrainer, en cas de non intervention, le denoyage du cur du reacteur, sa montee en temperature, la fonte des crayons combustibles et des structures qui les maintiennent. On peut alors aboutir a la degradation complete du cur et au percement de la cuve par les debris fondus (le corium). Le corium a haute temperature (2000 a 3000 k) peut ainsi couler sur le radier en beton du reacteur et l'eroder rapidement, comme l'ont montre plusieurs programmes experimentaux. De cette interaction, on a observe, entre autre, le degagement d'un epais nuage d'aerosols et d'importants flux de chaleur. L'effet de ces aerosols sur la propagation du flux de chaleur emis par le bain de corium a ete mis en evidence au cours de ce travail. Nous avons ecrit un modele numerique de transfert radiatif dans un milieu capable d'absorber, de diffuser et d'emettre de l'energie. Des resultats experimentaux puises dans la litterature nous ont permis de degager des elements de validation de ce modele et de montrer clairement l'effet d'ecran lie aux aerosols. A partir de ce modele, nous avons ensuite etabli des correlations relatives a des essais particuliers (essais l1, l2, l4 et l7 du programme advanced containment experiment). Ces correlations donnent l'extinction moyenne due aux aerosols en fonction de la concentration moyenne en aerosols dans le milieu. Elles sont destinees a etre ajoutees aux logiciels d'analyse de l'interaction corium/beton lesquels, en majorite, ne tiennent pas compte de la presence des aerosols et surestiment les pertes radiatives vers le haut de l'enceinte. Nous avons applique l'une de ces correlations a l'essai l7 a l'aide du logiciel corcon-uw. Nous avons ainsi montre que la prise en compte des aerosols rapproche significativement nos calculs des resultats experimentaux et nous permet d'observer le blocage de l'energie thermique pres du bain et l'elevation de la temperature du corium qui en resulte

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Plevacova, Kamila. "Etude des matériaux sacrificiels absorbants et diluants pour le contrôle de la réactivité dans le cas d'un accident hypothétique de fusion du coeur de réacteurs de quatrième génération." Phd thesis, Université d'Orléans, 2010. http://tel.archives-ouvertes.fr/tel-00592463.

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Afin de limiter les conséquences d'un hypothétique accident grave avec la fusion du coeur dans un réacteur à neutrons rapides de génération IV refroidi au sodium, la recriticité doit être évitée au sein du mélange de combustible oxyde et de structures fondus, appelé corium. Pour cela, des matériaux absorbants, tels que le carbure de bore B4C, seront utilisés dans ou près du coeur, et des matériaux diluants dans le récupérateur de corium. L'objectif de ce travail est de présélectionner des matériaux parmi ces deux types de familles et de comprendre leur comportement au contact avec le corium. Concernant le B4C, des calculs thermodynamiques et des expériences ont permis de conclure à la formation de deux phases immiscibles dans le système UO2 - B4C à haute température, une oxyde et une borure, ainsi qu'à la volatilisation d'une partie de l'élément absorbant bore. Cette séparation de phases pourra réduire l'efficacité de l'absorption neutronique au sein de la phase oxyde. Une solution à ce comportement serait d'augmenter la quantité de B4C ou d'utiliser un absorbant oxyde miscible avec le combustible. Eu2O3 ou HfO2 pourraient convenir car il a été montré qu'ils forment une solution solide avec UO2. Concernant le matériau diluant, les oxydes mixtes Al2O3 - HfO2 et Al2O3 - Eu2O3 ont été étudiés. L'interaction de ces systèmes avec UO2 étant inconnue à ce jour, les premiers points ont été recherchés sur les diagrammes ternaires correspondants. Contrairement au système Al2O3 - Eu2O3 - UO2, le mélange Al2O3 - HfO2 - UO2 présente un seul eutectique et donc un seul chemin de solidification ce qui permet de prévoir plus facilement la manière dont le corium solidifierait dans le récupérateur.

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Sanchez-Brusset, Mathieu. "Mécanismes d'oxydation de l'acier liquide lors de l'Interaction Corium-Béton à haute température en cas d'accident grave de réacteur nucléaire." Thesis, Perpignan, 2015. http://www.theses.fr/2015PERP0015/document.

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En cas d' accident grave de réacteur nucléaire, la perte de réfrigérant peut conduire à la formation d'un mélange liquide à haute température (T>2500K) constitué majoritairement du combustible nucléaire et des matériaux de structure (corium). En cas de rupture de la cuve, le corium est susceptible d'interagir avec le béton de l'enceinte de confinement. Au contact du béton, la présence d'acier liquide modifie les processus d'ablation du béton et entraine une production de H2 et CO. Les objectifs de cette thèse étaient de déterminer la cinétique d'oxydation de l'acier liquide dans ces conditions, et d'identifier les mécanismes prépondérants. Pour répondre à ces objectifs, trois volets ont été développés: une approche à l'équilibre thermodynamique, des expériences analytiques à effets séparés et des expériences intégrales avec du corium prototypique. L'analyse des expériences intégrales montre que les gaz relâchés par le béton ne sont pas les seules sources d'oxydation, mais qu'une source d'oxydation extérieure au béton participe aux mécanismes d'oxydation. Les expériences analytiques ainsi que les calculs à l'équilibre thermodynamique ont montré que le corium, par sa capacité à devenir sur-stoechiométrique, est une source d'oxydation supplémentaire. Au contraire, les oxydes du béton ne participent pas au mécanisme d'oxydation. Le mécanisme d'oxydation de l'acier liquide est basé sur une oxydation relativement forte du chrome et du fer. Le nickel n'est pas oxydé, et serait consommé préférentiellement par Évaporation d'après les calculs thermodynamiques. L'étude cinétique de l'oxydation a permis d'une part d'établir deux lois cinétiques d'oxydation par O2 et CO2 et d'autre part de proposer une modélisation de la cinétique d'oxydation de l'acier lors des essais intégraux
In case of severe nuclear accident, the loss of coolant leads to the formation of a high temperature liquid mixture (T>2500K) of nuclear fuel and structural materials inside the vessel. After the vessel failure, the corium could interact with the concrete of the reactor pit. The metallic phase inside the corium during corium-concrete interaction, changes the ablation processes and release H2 and CO. The aim of the PhD thesis was to study the kinetics and mechanisms of the liquid steel oxidation during corium-concrete interaction. In this way, the study was divided in three parts: with calculations at the thermodynamic equilibrium, with analytical experiments and with prototypical experiments. The results of oxidation analyses during prototypical experiments show that gases inside the concrete are not the only one source of oxidation and that another source outside the concrete have to participate to the oxidation mechanism. The analytical experiments and the thermodynamic approach show that the corium can oxidize the metallic phase whereas the concrete oxides cannot. The oxidation mechanism of liquid steel is based on high chromium and iron oxidation leading to their depletion. Oxidation of nickel does not occur, it would be mainly evaporated according to the thermodynamic calculations. Thanks to the kinetic study, the rates of the liquid steel oxidation by O2 et CO2 have been found and a phenomenological model have been proposed to estimate the steel oxidation during the prototypical experiments

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Villarreal, Larrauri Alejandro. "Analysis and modeling of ex-vessel underwater cooling processes of debris bed and molten corium pool in interaction with concrete." Electronic Thesis or Diss., Université de Lorraine, 2020. http://www.theses.fr/2020LORR0022.

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En cas d'accident grave avec fusion du cœur, le magma surchauffé constitué d'acier et de combustible fondu, appelé corium (T> 2 500 K), peut menacer l'intégrité de la cuve du réacteur, puis du bâtiment de confinement, si le refroidissement du corium n’est pas assuré. La capacité de refroidissement en situation hors-cuve, par l’injection d’eau et pénétration de celle-ci dans le corium en surface supérieure, est étudiée pour deux configurations attendues : le lit de particules et bain de corium. La seconde configuration est liée à la situation d’interaction corium-béton (ICB) où une croûte se forme en face supérieure en contact avec l’eau, puis est soumise à une fracturation à cause des effets thermiques dans cette croûte. L’enjeu est de caractériser l’efficacité d’une éventuelle pénétration de l’eau dans la croûte. La première configuration peut intervenir en particulier dans deux situations suite à une fragmentation du corium dans l’eau : lors de l’éjection hors de la cuve, ou suite à des périodes d’éjection à travers la croute en phase d’ICB par entraînement du corium par des gaz issus de l’ablation du béton. Les phénomènes de pénétration de l’eau dans le corium sont examinés par une analyse approfondie des résultats des expériences disponibles, par la mise au point d’un modèle analytique 1D et par la modification et l’utilisation du code de thermohydraulique multiphasique multi-fluides (CMFD) MC3D. L’analyse 1D permet de mieux comprendre les détails de l’écoulement diphasique dans la matrice poreuse et conduit à proposer un modèle simplifié de pénétration de l’eau, avec des relations correspondantes applicables pour les deux configurations d'intérêt. Par ailleurs, le développement et l’impact d’instabilités au front de pénétration sont étudiés avec des simulations 2D avec MC3D, illustrant le rôle important de la température initiale du lit et sa perméabilité sur la vitesse de pénétration du front, et sur l’apparition des instabilités. Le modèle analytique est alors étendu à une configuration à deux zones (une zone soumise à un écoulement diphasique en contre-courant et une zone monophasique dans laquelle la vapeur surchauffée traverse) pour analyser plus en détail l’impact des hétérogénéités de progression du front sur les flux thermiques extraits. Le mécanisme de pénétration de l’eau dans les croûtes est discuté. L’analyse indique de forts effets de bords sur les processus de fracturation dans les essais SSWICS (Argonne Nat. Lab.), dédiés à ce phénomène. Les conclusions des travaux précédents sur l’efficacité du phénomène ne peuvent dès lors être confirmées du fait des fortes incertitudes sur les processus de fracturation, très sensibles aux propriétés mécaniques du corium, très mal caractérisées. Finalement, les modèles sont appliqués aux situations réelles impliquant la présence de la puissance résiduelle. Pour les lits de débris, les flux extraits et les capacités de refroidissement sont moindres qu’avec l’utilisation du critère simplifié de « flux d’assèchement »
In the case of a hypothetical nuclear severe accident with partial or extensive core meltdown, the superheated magma made of molten steel and fuel, called corium (T > 2500K), may threaten the integrity of the reactor pressure vessel and subsequently the reactor containment building, if long-term corium coolability is not assured. The coolability by water injection and successive water penetration into the corium through the upper surface is analyzed for two expected configurations: particle bed, and corium pool overlaying the concrete. The second configuration is linked to the situation of Molten Corium-Concrete Interaction (MCCI), where a crust is formed in the upper corium surface when it comes into contact with water and is later subjected to thermal stresses that lead to its fracturing. The challenge is to characterize the effectiveness of extracting heat by the possible water penetration into the crust. The first configuration can be expected in two different situations: melt fragmentation coming from the rupture of the reactor pressure vessel and expulsion of the corium, or during melt eruption episodes through the corium crust during MCCI via corium entrainment by the concrete decomposition gases. The phenomena linked to the water penetration into the corium for these two configurations are examined through an in-depth analysis of the available experimental results, by the development of an analytical model and finally through the modification and use of the Computational Multi-Fluid Dynamics (CMFD) code MC3D. One dimensional analysis conducts to a better understanding of the minutia of the two-phase countercurrent flow through the porous media and leads the proposal of a simplified heat flux model for the water penetration with corresponding relations applicable for both configurations of interest. Furthermore, the development and the impact of penetrating front instability are studied with the help of 2D MC3D simulations, which show important effects of the initial temperature and the permeability of the corium configuration on the penetration front velocity and appearance of the instabilities. The analytical model is extended to a pseudo-two-dimensional two-zone configuration (with one zone subjected to a two-phase countercurrent flow and another through which monophasic superheated vapor flows) to analyze in greater detail the impact of the penetrating front heterogeneity over the extracted heat flux. The mechanism of water penetration through a fractured crust is revisited. The analysis indicates strong border effects in the SSWICS tests (Argonne National Laboratories) dedicated to the study of this phenomenon. The conclusions of precedent studies on the efficiency of the phenomena could not, therefore, be confirmed due to important uncertainties over the process of fracturing, overly sensitive to the mechanical properties of corium, which in turn are not properly characterized. Finally, the models, and simulations, are applied to real accidental scenarios, including the presence of residual power. For the debris bed, the extracted heat flux, and the cooling capabilities are less than those found using the simplified dry-out heat flux criteria

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Mastori, Helena. "Mécanismes de dégradation des bétons lors de l'interaction corium-béton." Thesis, Aix-Marseille, 2019. http://www.theses.fr/2019AIXM0069.

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Cette thèse porte sur la caractérisation de bétons siliceux (S) et silico-calcaires (SC) lorsqu’ils sont exposés à des températures élevées. Une hypothèse motivant ce travail est que la compréhension de la dégradation des propriétés de ces bétons, en avance du front de fusion, pourrait apporter de nouvelles pistes pour interpréter les résultats de ces expériences. Des échantillons n’ayant jamais été mis en contact avec des métaux/oxydes en fusion ont d’abord été étudiés. La thermogravimétrie, la porosimétrie par intrusion de mercure et l’impédancemétrie complexe ont été utilisées pour décrire leurs propriétés après qu’ils ont été soumis à des températures pouvant atteindre 1000°C. Les résultats de l'ATG ont permis l’identification de domaines de température dans lesquels des mécanismes de dégradation spécifiques sont activés. Ceux de porosimétrie ont montré que les volumes poreux et la taille typique des pores augmentent de manière importante avec la température. Il est par ailleurs démontré qu’à 1000°C, la surface d’échange des bétons SC est deux fois plus importante que celle des bétons S. Enfin, les tortuosités élevées obtenues par impédancemétrie suggèrent une topologie des réseaux poreux d’une grande complexité. Dans une deuxième partie de cette thèse, les échantillons de bétons étudiés ont été préalablement mis en contact avec des métaux et/ou des oxydes en fusion. Ils ont été analysés par tomographie X ou par microscopie électronique à balayage. Aucun phénomène d’imprégnation des phases métal/oxyde n’a pu être observé. Des signatures de possibles phénomènes de percolation de ces phases par des mécanismes de décarbonatation ont toutefois été mises en évidence
This thesis deals with the characterization of siliceous (S) and limestone-siliceous (SC) concretes when exposed to high temperatures. The understanding of the degradation of their properties, in advance of the melt front, is the hypothesis that motivates this work since it could bring new avenues to interpret the results of these experiments. Samples that have never been in contact with molten metals/oxides were first studied. Thermo-gravimetry, mercury intrusion porosimetry and complex impedancemetry were used to describe their properties after they were subjected to temperatures up to 1000°C. Thermo-gravimetric analyses allowed the identification of temperature domains in which specific degradation mechanisms are activated. Those of porosimetry showed that porous volumes and typical pore sizes increase significantly with the temperature. It is also demonstrated that at 1000°C, the exchange surface of SC concretes is twice as large as that of Sconcretes. Finally, the high tortuosity obtained by impedancemetry suggests a topology of porous networks of great complexity. In a second part of this thesis, the studied concrete samples were previously in contact with molten metals and/or oxides. They were analysed by X-ray tomography or scanning electron microscopy. No phenomenon of impregnation of the metal/oxide phases could be observed. Signatures of possible phenomena of percolation of these phases by decarbonation mechanisms have however been demonstrated

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Le, Roy de Bonneville Florian. "Modélisation numérique de l'agitation et du mélange dans les écoulements à bulles. Application aux phénomènes de convection dans un bain de corium." Thesis, Toulouse, INPT, 2020. http://www.theses.fr/2020INPT0088.

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Les écoulements à bulles font partie de la famille des écoulements polyphasiques dans lesquels des particules, solides, liquides ou gazeuses, sont dispersées dans un fluide porteur. Ce type d’écoulements est très courant, on le retrouve dans nombreux procédés industriels (colonnes à bulles, colonnes d’extraction, lits fluidisés, décanteurs) et naturels (vagues déferlantes, panaches volcaniques). La présence des bulles joue notamment un rôle majeur dans les accidents nucléaires de fusion du coeur en influençant la dynamique du bain de corium. Cette présence dans des do-maines très variés a favorisé un développement important de méthodes expérimentales et numériques pour étudier ce type d’écoulements. Dans cette étude, nous nous intéressons à l’écoulement induit par l’ascension d’un essaim de bulles millimétriques (dont le nombre de Reynolds est de plusieurs centaines) dans un liquide. Dans cette situation, les interactions entre les sillages jouent un rôle majeur conduisant à une agitation turbulente aux caractéristiques originales. L’une des plus frappantes est l’existence d’un régime spectral singulier où l’énergie des fluctuations des vitesse du liquide évolue en puissance -3 du nombre d’onde. Fondamentalement, nous souhaitons comprendre les mécanismes de transfert turbulent interéchelle afin de modéliser les processus de mélange et de transfert dans les applications. Pour cela nous proposons de simuler l’écoulement en couplant une description eulérienne de la phase porteuse à une méthode Lagrangienne pour le suivi des bulles. Dans notre approche numérique, l’action de chaque bulle sur le liquide est modélisée par une source volumique de quantité de mouvement répartie sur quelques éléments de maillage. Les plus petites échelles de l’écoulement (c’est-à-dire des échelles beaucoup plus petites que le diamètre des bulles) ne sont pas finement résolues. Ce choix de nous concentrer sur les grandes échelles de l’écoulement nous permet de simuler des fractions volumiques conséquentes avec un grand nombre de bulles avec une puissance de calcul raisonnable. Pour calculer la trajectoire des bulles, nous utilisons les forces hydrodynamiques que le liquide exerce sur chacune d’elles. Ceci nécessite de déterminer la perturbation qu’une bulle crée dans son voisinage afin d’annuler la force que la bulle exerce artificiellement sur elle-même. Nous avons établi un modèle pour déterminer cette perturbation nous permettant ainsi de calculer de façon précise les forces de traînée et de masse ajoutée. Grâce à cette méthode, nous avons simulé l’agitation induite par l’ascension d’un essaim de bulles homogène et obtenu des résultats en bon accord avec l’expérience. Une fois validées, ces simulations permettent d’étudier le bilan entre production, dissipation et transfert inter-échelle dans le plan spectral pour analyser les mécanismes de la turbulence induite par les bulles. Dans un but de prévention des risques, le modèle numérique est ensuite appliqué à la simulation d’un bain de corium produit lors d’un accident de fusion du cœur d’une centrale nucléaire. La dynamique d’ablation du béton est directement liée à la répartition des flux de chaleur aux parois qui mettent principalement en jeu les phénomènes de convection turbulente d’origine thermique et ceux induits par les bulles
Bubbly flows belong to the family of multiphase flows in which particles, whether solid, liquidor gaseous, are dispersed in a carrier fluid. This type of flow is very common and can be found inmany industrial processes (bubble columns, extraction columns, fluidized beds, decanters) and natural processes (breaking waves, volcanic plumes). In particular, the presence of bubbles plays a major role in nuclear core meltdown accidents by influencing the dynamics of the corium bath.This presence in a wide variety of fields has led to the significant development of experimental and numerical methods to study this type of flow. In this study, we are interested in the flow induced by the rise of a swarm of millimetre-sizedbubbles (with a Reynolds number of several hundred) in a liquid. In this situation, interactions between the wakes play a major role leading to turbulent agitation with original characteristics. One of the most striking is the existence of a singular spectral regime where the energy of the fluctuations in the liquid velocity evolves in power -3 of the wave number. Fundamentally, we are in-terested in understanding the interscale turbulent transfer mechanisms in order to model mixing and transfer processes in applications. For this purpose we propose to simulate the flow by coupling an Eulerian description of the carrier phase to a Lagrangian method for the bubbles. In our numerical approach, the action of each bubble on the liquid is modelled by a volume source of momentum distributed over a few mesh elements. The smallest scales of the flow (i.e. scales much smaller than the bubble diameter) are not finely resolved. This choice to focus on the large scales of the flow allows us to simulate large volume fractions with a large number of bubbles with reasonable computing resources. To calculate the trajectory of the bubbles, we use the hydrodynamic forces that the liquid exerts on each of them. This requires us to determine the perturbation that a bubble creates in its vicinity in order to cancel the force that the bubble artificially exerts on itself. We have developed a model to determine this perturbation allowing us to accurately calculate the drag and added-mass forces. Using this method, we simulated the agitation induced by the rise of a homogeneous swarm of bubbles and obtained results in good agreement with the experiment. Once validated, these simulations allow us to study the budget between production, dissipation and inter-scale transfer in the spectral domain to analyze the mechanisms of bubble-induced turbulence. For risk prevention purposes, the numerical model is then applied to the simulation of a corium bath produced during a core meltdown accident in a nuclear power plant. The dynamics of concrete ablation is directly related to the distribution of heat fluxes to the walls, which mainly involve turbulent convection phenomena of thermal origin and those induced by bubbles

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Amižić, Milan. "Interaction corium-béton : étude du transfert de chaleur en écoulement diphasique." Thesis, Grenoble, 2014. http://www.theses.fr/2014GRENI002.

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Dans le cadre de la recherche sur les accidents graves pour la deuxième et la troisième génération de réacteurs nucléaires, certains aspects de l'ablation de béton dans le puits de cuve au cours de l'interaction corium-béton (ICB) restent encore inexpliquées. La détermination d'échange de chaleur le long de la région interfaciale entre un bain de corium et un béton est importante pour l'évaluation de la progression d'ablation du béton et, éventuellement, la percée de fondation. Le projet CLARA s'inscrit une recherche expérimentale sur la thermohydraulique au sein d'un bain de liquide agitée par des bulles de gaz. Les essais CLARA sont réalisés avec des matériaux simulants. Ils permettent de mettre en évidence l'influence de la vitesse superficielle du gaz, de la viscosité du liquide et de la géométrie sur le coefficient d'échange de chaleur entre le bain de liquide chauffé et les parois verticales et horizontales de la piscine qui sont maintenues à une température uniforme. La première campagne d'essais a été réalisée avec la configuration du bain de petite taille (50 cm × 25 cm × 25 cm). Les essais ont été réalisés avec des liquides couvrant un large éventail de viscosité dynamique, d'environ 1 mPa s à 10000 mPa s. La vitesse superficielle du gaz est modifiée jusqu'à 8 cm/s. Cette thèse comporte une brève description de la phénoménologie de l'ICB, une synthèse bibliographique sur les corrélations d'échange de chaleur existantes pour l'écoulement diphasique et le taux de vide, une description de l'installation CLARA, les résultats des essais et leur interprétation. Les résultats expérimentaux sont comparés avec les modèles existants et certains nouveaux modèles pour l'évaluation du coefficient d'échange de chaleur dans un écoulement diphasique
In the context of severe accident research for the second and the third generation of nuclear power plants, there are still open issues concerning some aspects of the concrete cavity ablation during the molten corium - concrete interaction (MCCI). The determination of heat transfer along the interfacial region between the molten corium pool and the ablating basemat concrete is crucial for the assessment of concrete ablation progression and eventually the basemat meltthrough. For the purpose of experimental investigation of thermalhydraulics inside a liquid pool agitated by gas bubbles, the CLARA project has been launched. The CLARA experiments are performed using simulant materials and they reveal the influence of superficial gas velocity, liquid viscosity and pool geometry on the heat transfer coefficient between the internally heated liquid pool and vertical and horizontal pool walls maintained at uniform temperature. The first test campaign has been conducted with the small pool configuration (50 cm × 25 cm × 25 cm). The tests have been performed with liquids covering a wide range of dynamic viscosity from approximately 1 mPa s to 10000 mPa s and the superficial gas velocity is varied up to 8 cm/s. This thesis comprises a brief description of MCCI phenomenology, literature reviews on the existing heat transfer correlations for twophase flow and the void fraction, a description of CLARA setup, experimental results and their interpretation. The experimental results are compared with existing models and some new models for the assessment of heat transfer coefficient in two-phase flow
U kontekstu istraživanja teških nesre´ca u nuklearnim elektranamadruge i tre´ce generacije, neka pitanja vezana za ablaciju temelja kontejnmentatijekom interakcije rastaljenog korijuma i betona i dalje ostajuotvorena. Odred¯ivanje prijenosa topline u površinskom podrucˇjuizmed¯u bazena rastaljenog korijuma i betona kljucˇno je za odred¯ivanjenapredovanja ablacije i u konaˇcnici procjene vremena rastapanjatemelja kontejnmenta. U svrhu eksperimentalnog istraživanja prijenosatopline u tek´cinama miješanima ubrizgavanjem zraka, pokrenutje projekt nazvan CLARA.CLARA eksperimenti izvode se koriste´ci imitacijske materijale i otkrivajuutjecaj fiktivne brzine plina, viskoznosti teku´cine i geometrijebazena na koeficijent prijenosa topline izmed¯u grijanog bazena te njegovihvetrikalnih i horizontalnih stijenki ˇcija se temperatura održavana konstantnoj temperaturi. Prva serija eksperimenata provedena je sbazenom male konfiguracije (50 cm × 25 cm × 25 cm). Eksperimentisu izvedeni s teku´cinama dinamiˇcke viskoznosti od približno 1 mPas do 10000 mPa s, dok je maksimalna fiktivna brzina plina 8 cm/s.Ova disertacija sadrži kratak opis fenomenologije procesa interakcijerastaljenog korijuma i betona, pregled postoje´cih korelacija zaviprijenos topline u dvofaznom toku i korelacija za poroznost, opisCLARA eksperimentalne postave, rezultate eksperimenta i njihovuinterpretaciju. Rezultati eksperimenta su uspored¯eni s predvid¯anjimaprema postojec´im modelima. Predloženi su takod¯er i neke nove korelacijeza odred¯ivanje koeficijenta prijenosa topline u dvofaznom toku

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BATTAIL, CLARET SYLVIE. "Accident severe dans les reacteurs a eau pressurisee : interaction corium-eau." Paris 11, 1993. http://www.theses.fr/1993PA112263.

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Pour etudier le phenomene d'interaction entre le corium et l'eau, on propose un scenario pour decrire le comportement d'une goutte d'oxyde de fer fondu brusquement plongee dans un bain de liquide a temperature ambiante. En premier lieu, on s'est interesse plus en detail a la modelisation de l'evolution du film de vapeur qui entoure la goutte chaude comprenant la phase d'etablissement d'un film stable et la phase de destabilisation de ce film au passage d'une onde de pression externe. Par ailleurs, on a modelise le processus de fragmentation du corps chaud induit par la destabilisation par un processus du a l'impact de micro-jets d'eau liquide avec piegeage d'eau dans le corps chaud. Enfin, un modele dit de dynamique de bulle a ete propose pour decrire l'evolution de la bulle vapeur alimentee par les fragments. Les resultats theoriques ainsi obtenus sont compares a des resultats experimentaux

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Namiech, Julien. "Fragmentation d'un jet de corium lors de sa chute dans l'eau." Grenoble INPG, 2002. http://www.theses.fr/2002INPG0043.

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Durant un accident grave de réacteur nucléaire à eau sous pression une large masse de corium pourrait couler sous la forme d'un jet compact en fond de cuve. Au contact de l'eau, restée en fond de cuve, ce jet subirait une fragmentation intense qui pourrait conduire à un prémélange important de corium et d'eau susceptible de donner lieu à une explosion de vapeur, capable de menacer l'intégrité de la cuve. Afin de quantifier la fragmentation de ce jet de corium, une étude analytique a été développée. Cette étude consiste principalement à modéliser l'écoulement de vapeur autour du jet et l'instabilité qui se développe à sa surface. Par rapport aux études précédentes, une attention plus particulière est portée sur les particules issues de la fragmentation du jet et qui interagissent avec l'écoulement de vapeur. Un modèle complet est élaboré afin de calculer dans chaque situation, caractérisée par des conditions initiales, la longueur de rupture du jet et le diamètre des particules éjectées. Ce modèle s'appuie principalement sur des résultats de couches limites et sur des calculs de stabilité linéaire. Les résultats de ce modèle complet sont comparés aux principales expériences de ce domaine et une corrélation finale des résultats est établie. L'accord sur la longueur de rupture est satisfaisant, cependant le diamètre prédit pour les particules tend à être trop élevé. Ce dernier résultat pourrait s'expliquer par une fragmentation secondaire des particules dans l'eau et par une incertitude importante sur l'écoulement de vapeur.

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Books on the topic "Corium"

M, Ishii, U.S. Nuclear Regulatory Commission. Division of Systems Analysis and Regulatory Effectiveness., and Purdue University. School of Engineering., eds. Corium dispersion in direct containment heating. Washington, DC: Division of Systems Analysis and Regulatory Effectiveness, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, 1999.

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R, Greene Sherrell, U.S. Nuclear Regulatory Commission. Division of Safety Issue Resolution., and Oak Ridge National Laboratory, eds. BWR Mark II ex-vessel corium interaction analyses. Washington, DC: Division of Safety Issue Resolution, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, 1991.

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W, Spencer B., U.S. Nuclear Regulatory Commission. Office of Nuclear Regulatory Research. Division of Systems Research., and Argonne National Laboratory, eds. Fragmentation and quench behavior of corium melt streams in water. Washington, DC: Division of Systems Research, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, 1994.

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O'Leary, David. Differences in strength between the grain and corium layers of bovine leather. Northampton: Nene College, 1996.

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M, Ishii, U.S. Nuclear Regulatory Commission. Office of Nuclear Regulatory Research. Division of Systems Research., and Purdue University. School of Nuclear Engineering., eds. Air-water simulation of phenomena of corium dispersion in direct containment heating. Washington, DC: Division of Systems Research, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, 1994.

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T, Haalstra R., and Peterse D. J, eds. Cattle footcare and claw trimming: The origin and prevention of the necrotising inflammations of the corium (ulcerations of the claw). 3rd ed. Ipswich, U.K: Farming Press Books, 1989.

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Moorcock, Michael. Corum. London: Millennium, 1992.

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Book chapters on the topic "Corium"

Bährle-Rapp, Marina. "Corium." In Springer Lexikon Kosmetik und Körperpflege, 129. Berlin, Heidelberg: Springer Berlin Heidelberg, 2007. http://dx.doi.org/10.1007/978-3-540-71095-0_2423.

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Gabrys, Beata, John L. Capinera, Jesusa C. Legaspi, Benjamin C. Legaspi, Lewis S. Long, John L. Capinera, Jamie Ellis, et al. "Corium." In Encyclopedia of Entomology, 1063. Dordrecht: Springer Netherlands, 2008. http://dx.doi.org/10.1007/978-1-4020-6359-6_10030.

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Mehlhorn, Heinz. "Corium." In Encyclopedia of Parasitology, 586. Berlin, Heidelberg: Springer Berlin Heidelberg, 2016. http://dx.doi.org/10.1007/978-3-662-43978-4_706.

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Mehlhorn, Heinz. "Corium." In Encyclopedia of Parasitology, 1. Berlin, Heidelberg: Springer Berlin Heidelberg, 2015. http://dx.doi.org/10.1007/978-3-642-27769-6_706-2.

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Maity, Ram Kumar, T. Sundararajan, M. Rajendrakumar, and K. Natesan. "Toward Analysis of Corium Hydraulics in Liquid Sodium." In Lecture Notes in Mechanical Engineering, 595–608. Singapore: Springer Nature Singapore, 2024. http://dx.doi.org/10.1007/978-981-99-7827-4_47.

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Wei, Wei, and Xin-rong Cao. "The Simulation of Corium Dispersion in Direct Containment Heating Accidents." In Zero-Carbon Energy Kyoto 2009, 274–78. Tokyo: Springer Japan, 2010. http://dx.doi.org/10.1007/978-4-431-99779-5_43.

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Kumar, Nirmal, Varun Mishra, D. Faisal, R. K. Chaudhary, V. Chaudhry, and S. M. Ingole. "Structural Integrity Assessment of Calandria End-Shield Assembly for In-Vessel Corium Retention Under Severe Accident Condition." In Lecture Notes in Mechanical Engineering, 415–33. Singapore: Springer Singapore, 2022. http://dx.doi.org/10.1007/978-981-16-8724-2_38.

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Bär, B. W., and W. Tausch. "Vergleichende Untersuchungen zur Bandplastik bei der chronischen fibularen Bandinsuffizienz mit Corium, Faszie, PDS-Band und der Methode nach Watson-Jones." In Fortschritte in der Unfallchirurgie, 337–42. Berlin, Heidelberg: Springer Berlin Heidelberg, 1992. http://dx.doi.org/10.1007/978-3-642-77401-0_53.

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Krogh, Peter Gall, and Ilpo Koskinen. "Corpus." In Design Research Foundations, 123–27. Cham: Springer International Publishing, 2020. http://dx.doi.org/10.1007/978-3-030-37896-7_10.

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Gilmour, Jess K. "Corvus." In The Practical Astronomer’s Deep-sky Companion, 49–50. London: Springer London, 2003. http://dx.doi.org/10.1007/978-1-4471-0071-3_16.

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Conference papers on the topic "Corium"

Journeau, C., and J. M. Seiler. "Interaction Corium-Béton." In Premières conséquences du REX de Fukushima sur l’exploitation des réacteurs et installations nucléaires. Les Ulis, France: EDP Sciences, 2012. http://dx.doi.org/10.1051/jtsfen/2012pre14.

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Atkhen, K., M. Cranga, and Ch Journeau. "Interaction Corium-Béton." In Thermohydraulique des accidents graves dans les réacteurs à eau légère. Les Ulis, France: EDP Sciences, 2012. http://dx.doi.org/10.1051/jtsfen/2012the08.

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Frolov, Kyrill N., Michel Duclot, and Christophe Journeau. "Interface Stability Criteria for Ex-Vessel Corium Solidification." In 12th International Conference on Nuclear Engineering. ASMEDC, 2004. http://dx.doi.org/10.1115/icone12-49529.

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In case of a severe accident at a nuclear power plant (NPP) involving the reactor core melt-down and the subsequent reactor pressure vessel (RPV) melt-through, confident solidification of ex-vessel corium is considered to be the imperative condition of safe retention of corium within the plant containment in the long term. The rate-determining process for solidification of ex-vessel coriums in the long-term is the chemical diffusion in the liquid phase at the solid-liquid interface. The process of chemical diffusion in the diffusive boundary layer can evolve and take on different rates, depending on the boundary conditions and the melt composition. The chemical diffusion coefficient models presented to date in the literature resort to correlations of the former to the self-diffusion coefficients among other intrinsic properties. The general feature of such models is that they predict in the tracer limit that the main diagonal coefficients tend towards the self-diffusion coefficients whereas the cross-over terms cancel out. It is revealed in this study that this particular feature is characteristic of prototypic corium melts, mixtures of several major as well as minor components. Following the corium-concrete interaction, the multicomponent ex-vessel corium melts would contain certain fractions of silica. Accordingly, they are considered in this paper as silicate oxide melts. As a first contribution, this paper comes up with a development of interface stability criteria for its application to solidification of silicate oxide melts. For ex-vessel corium melts, not far removed from equilibrium solidification conditions (long-term retention), the extension of the FICK’s law due to ONSAGER can be applied to description of the diffusive mass transfer. This formalism implies that the diffusive fluxes are linear combinations of the products of the phenomenological diffusion coefficients and the driving forces (gradients). In this regard, the near equilibrium solidification of corium containing N components is determined by a (N−1)(N−1) chemical diffusion matrix comprising the proper and cross-over coefficients. On the other hand, by comparison to the constitutional super cooling criterion of the interface stability in the course of solidification, a relationship between the macroscopic solidification conditions and the phenomenological coefficients can be established. This analysis, earlier developed for liquid metal solidification, has received attention in this study in view of its extension onto the solidification of multicomponent oxide melts. The important conclusion whatsoever is that the crossover terms can’t be neglected as easily as in the case of liquid metal alloys, particularly for silicate melts. As in prototypic corium compositions certain tracer diffusion coefficients can be relatively easily obtained from an experiment, this paper suggests an empirical method for determination of corium constituents’ self-diffusivities from controlled solidification tests.

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Xu, Zhichun, Yapei Zhang, G. H. Su, Wenxi Tian, and Suizheng Qiu. "Numerical Simulation of Concrete Ablation and Corium Cooling for Molten Corium-Concrete Interaction (MCCI)." In 2020 International Conference on Nuclear Engineering collocated with the ASME 2020 Power Conference. American Society of Mechanical Engineers, 2020. http://dx.doi.org/10.1115/icone2020-16388.

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Abstract In a postulated severe accident situation in Light Water Reactors (LWRs), if the core fuel cannot be effectively cooled, the reactor core material will be heated and form a molten corium in the lower head. When the lower plenum of the reactor vessel fails, the molten corium may flow into the cavity under the reactor vessel and react with the concrete. This process, known as Molten Corium Concrete Interaction (MCCI), is characterized by concrete ablation and oxidation of metal in the corium, both of which produce a large amount of combustible and non-condensable gases, threatening the integrity of the containment. Thus in-depth study of the characteristics of concrete ablation and corium cooling have great significance. In the present study, an MCCI analysis code, MOQUICO (molten corium concrete interaction and corium cooling code, QUI means quintic) has been developed. The MACE M3b and OECD/MCCI CCI-3 tests were analyzed to validate the developed code. The melt temperature, axial and radial ablation depths, upward heat flux were calculated and were in good agreement with the experimental measurements, which proved that the code is capable of simulating MCCI and related phenomena of LWRs. Sensitivity analyses on the factors of decay heat, concrete type and water injection moment were performed and analyzed.

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Journeau, Christophe, Viviane Bouyer, Nathalie Cassiaut-Louis, Pascal Fouquart, Pascal Piluso, Gérard Ducros, Stéphane Gossé, et al. "SAFEST Roadmap for Corium Experimental Research in Europe." In 2016 24th International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2016. http://dx.doi.org/10.1115/icone24-60916.

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SAFEST (Severe Accident Facilities for European Safety Targets) is a European project networking the European corium experimental laboratories with the objective to establish coordination activities, enabling the development of a common vision and research roadmaps for the next years, and of the management structure to achieve these goals. In this frame, a European roadmap on corium experimental research has been written to define research challenges to contribute to further reinforcement of Gen II and III NPP safety. It is based on the research priorities determined by SARNET SARP group as well as those from the recently formulated in the NUGENIA Roadmap for severe accidents and the recently published NUGENIA Global Vision report. It also takes into account issues identified in the analysis of the European stress tests and from the interpretation of the Fukushima accident. 19 relevant issues related to corium have been selected during these prioritization efforts. These issues have been compared to a survey of the European corium experimental facilities and corium analysis laboratories. Finally, the coherence between European infrastructures and R&D needs has been assessed and a table linking issues and infrastructures has been derived. It shows a few lacks in EU corium infrastructures, especially in the domains of core late reflooding impact on source term, Reactor Pressure Vessel failure and corium release, Spent Fuel Pool accidents, as well as the need for a large mass (100–500 kg) prototypic corium facility.

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Min, B. T., H. D. Kim, J. H. Kim, S. W. Hong, and I. K. Park. "Particle Size Characteristics of Molten Corium Quenched in Water." In 16th International Conference on Nuclear Engineering. ASMEDC, 2008. http://dx.doi.org/10.1115/icone16-48773.

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During a hypothetical severe accident in a nuclear reactor, a steam explosion might occur when molten corium interacts with water. The strength of a steam explosion affects the integrity of the containment of a nuclear reactor and is highly dependant on the characteristics of the melt-water-steam mixture. Since a break-up and fragmentation process during a pre-mixing are important mechanisms for a steam explosion behavior and affect the debris size distribution, the particle size characteristics of quenched corium have been investigated. For several years, series of experiments have been performed using prototypical corium in the TROI test facility with a high frequency induction heating using cold crucible technology. The molten corium was discharged into the cold water and the quenched debris particles were collected, sieved and examined for the effect of a size distribution on a steam explosion. The small corium droplets do not seem to contribute to a steam explosion owing to solidification at an early stage before the explosion but the large droplets contribute to it owing to their liquid state. It was also shown that single oxides and binary oxides with an eutectic composition (UO2/ZrO2 = 70/30 at weight percentage) led to steam explosions, but a binary oxide with a non-eutectic one did not. The mass mean diameters of the debris of the steam explosive composition was less than that of the non-steam explosive composition. Zirconia was the most energetic steam-explosive material in these tests, and an eutectic composition of corium also lead to a steam explosion, but a non-eutectic composition corium hardly led to a steam explosion. The particle sizes of the molten corium participating in a steam explosion were shown to be mainly 3–6 mm depending on the material and composition.

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Chen, Liang, Hua Pang, Ximing Xie, Lei Zhong, and Rong Cai. "Analysis of Transient Corium Pool Structure in the Lower Plenum of Reactor Vessel." In 2020 International Conference on Nuclear Engineering collocated with the ASME 2020 Power Conference. American Society of Mechanical Engineers, 2020. http://dx.doi.org/10.1115/icone2020-16470.

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Abstract The transient stratification of the corium in the lower plenum and its impact on the heat flux distribution on the outside of reactor vessel is analyzed in this work. A method for predicting the kinetic corium pool structure is proposed, which takes into account both thermo-chemical equilibrium and density evaluation of the corium. The transient stratification of the corium pool formed after a large loss of coolant accident (LLOCA) and a station blackout (SBO) accident of ACP1000 nuclear power plant in China was analyzed by this method. The transient structure of the corium pool was calculated at the moment when the amount of molten materials in the corium pool increased obviously. The results shown that the formation of a three-layer pool is highly possible when a two-layer pool is formed in the previous moment with a heavy metal layer on the bottom and the density of the heavy metal layer at the bottom is greater than the density of the newly added molten material at the next moment. The heat flux on the outside of the vessel wall faced the thin top metal layer and the vessel failure probability of the vessel here are high if a three-layer pool occurred.

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Louie, David L. Y., Yifeng Wang, Rekha Rao, Alec Kucala, and Jessica Kruichak. "Injectable Sacrificial Material System to Contain Ex-Vessel Molten Corium in Nuclear Accidents." In 2018 26th International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2018. http://dx.doi.org/10.1115/icone26-81440.

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An ongoing Sandia National Laboratories’ (SNL) research study is evaluating a potential design of an injectable sacrificial material (SM) system that could contain and cool corium ejected from a reactor vessel lower head failure during a potential severe accident involving melting fuel at a commercial light water nuclear reactor (LWR). An injectable system could be installed at any existing LWR, without significant modification to the cavity or to the drywell pedestal region of the plant. The conceptual design under consideration is a passive system. The SM is being optimized to quickly cool the corium mixture while creating gas to form porosity in the solid, such that subsequent water flooding can penetrate the structure and provide additional cooling. The SM would form a barrier and limit corium-concrete interactions. This three-year project takes a joint experimental and computational approach. In this paper, we will first discuss the success of our small-scale experiments conducted on the interactions between the surrogate corium material (SCM) and SM, used to evaluate the injectable concept. A larger experimental study, currently underway, will further validate the injectable concept, with a focus on accurately measuring interactions. This paper details the modeling study and its progress, including modeling the experiments on a surrogate system and extending the model to bench-scale corium flow from validation experiments. The project’s modeling studies will use the SNL engineering code suite SIERRA Mechanics to understand the interaction of injectable SM and molten corium and predict corium spreading. Spreading is modeled using a level set method to track the front in conjunction with a pressure-stabilized finite element method on the fully three-dimensional mass, momentum, and energy conservation equations. Using this diffuse-interface method, the corium spreading front can be tracked and an appropriate pseudo-solidification viscosity models can be implemented to accurately model the corium spreading physics. Finally, an injectable SM delivery system is discussed along with its deployment to the six-common commercial LWR designs currently operating in the United States. At the end of this project, a simplified model based on SIERRA simulations will be developed for implementation into MELCOR, a severe reactor analysis code, developed at SNL for the U.S. Nuclear Regulatory Commission. This will allow us to demonstrate the ability of the injectable SM system to mitigate the ex-vessel corium spreading, provide containment and negate the release of radionuclides.

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Johns, Jesse M., and Pavel V. Tsvetkov. "Autonomous Corium Reactor for Terrestrial Applications." In 2012 20th International Conference on Nuclear Engineering and the ASME 2012 Power Conference. ASME, 2012. http://dx.doi.org/10.1115/icone20-power2012-54922.

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Journeau, Christophe. "External Vessel Retention - Interaction corium - béton." In Les accidents graves. Les Ulis, France: EDP Sciences, 2017. http://dx.doi.org/10.1051/jtsfen/2017les08.

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Reports on the topic "Corium"

Huhtiniemi, I., H. Hohmann, and D. Magallon. FCI experiments in the corium/water system. Office of Scientific and Technical Information (OSTI), September 1995. http://dx.doi.org/10.2172/115058.

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Spencer, B. W., K. Wang, C. A. Blomquist, L. M. McUmber, and J. P. Schneider. Fragmentation and quench behavior of corium melt streams in water. Office of Scientific and Technical Information (OSTI), February 1994. http://dx.doi.org/10.2172/10136350.

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Spencer, B. W., J. J. Sienicki, and L. M. McUmber. Hydrodynamics and heat transfer aspects of corium-water interactions: Interim report. Office of Scientific and Technical Information (OSTI), March 1987. http://dx.doi.org/10.2172/6594223.

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Louie, David, Yifeng Wang, Rekha R. Rao, Alec Kucala, Kyle Ross, Jessica Nicole Kruichak, and William Robert Chavez. A New Method to Contain Molten Corium in Catastrophic Nuclear Reactor Accidents. Office of Scientific and Technical Information (OSTI), October 2019. http://dx.doi.org/10.2172/1573134.

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Magallon, D., and H. Hohmann. Experimental investigation of 150-KG-scale corium melt jet quenching in water. Office of Scientific and Technical Information (OSTI), September 1995. http://dx.doi.org/10.2172/115057.

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Farmer, M. T. The CORQUENCH Code for Modeling of Ex-Vessel Corium Coolability under Top FLooding Conditions. Office of Scientific and Technical Information (OSTI), August 2018. http://dx.doi.org/10.2172/1483840.

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W, Kim S., M. Z. Podowski, and R. T. Lahey. The modeling of core melting and in-vessel corium relocation in the APRIL code. Office of Scientific and Technical Information (OSTI), September 1995. http://dx.doi.org/10.2172/115069.

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Frid, W. Behavior of a corium jet in high pressure melt ejection from a reactor pressure vessel. Office of Scientific and Technical Information (OSTI), April 1988. http://dx.doi.org/10.2172/7013321.

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Morage, F., R. T. Lahey, Jr, and M. Z. Podowski. Experimental simulation of the water cooling of corium spread over the floor of a BWR containment. Office of Scientific and Technical Information (OSTI), September 1995. http://dx.doi.org/10.2172/115071.

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Sohal, Manohar Singh, and Larry James Siefken. A Heat Transfer Model for a Stratified Corium-metal Pool in the Lower Plenum of a Nuclear Reactor. Office of Scientific and Technical Information (OSTI), August 1999. http://dx.doi.org/10.2172/911498.

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