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Journal articles on the topic "U-Zr-Fe-O"

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Tsurikov, D. F., V. N. Zagryazkin, V. Yu Vishnevskii, E. K. Diakov, A. Yu Kotov, and V. M. Repnikov. "U–Zr–Fe–O Melt density." Atomic Energy 107, no. 4 (October 2009): 247–54. http://dx.doi.org/10.1007/s10512-010-9222-2.

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Asmolov, V. G., V. N. Zagryazkin, and D. F. Tsurikov. "The thermodynamics of U-Zr-Fe-O melts." High Temperature 45, no. 3 (June 2007): 305–12. http://dx.doi.org/10.1134/s0018151x07030042.

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Asmolov, V. G., V. N. Zagryazkin, and D. F. Tsurikov. "Estimation of the density of U-Zr-Fe-O melts." High Temperature 46, no. 4 (July 30, 2008): 579–82. http://dx.doi.org/10.1134/s0018151x08040202.

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Ohgi, Hiroshi, Yuji Nagae, and Masaki Kurata. "THERMODYNAMIC EVALUATION ON SOLIDIFICATION PATH FOR U-ZR-FE-O CORIUM." Proceedings of the International Topical Workshop on Fukushima Decommissioning Research 2022 (2022): 1066. http://dx.doi.org/10.1299/jsmefdr.2022.0_1066.

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Bottomley, Paul David W., Mairead Murray-Farthing, Dario Manara, Thierry Wiss, Bert Cremer, Cos Boshoven, Patrick Lajarge, and Vincenzo Rondinella. "Investigations of the melting behaviour of the U–Zr–Fe–O system." Journal of Nuclear Science and Technology 52, no. 10 (April 10, 2015): 1217–25. http://dx.doi.org/10.1080/00223131.2015.1023381.

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FUKASAWA, Masanori, Shigeyuki TAMURA, and Mitsuhiro HASEBE. "Development of Thermodynamic Database for U—Zr—Fe—O—B—C—FPs System." Journal of Nuclear Science and Technology 42, no. 8 (August 2005): 706–16. http://dx.doi.org/10.1080/18811248.2004.9726440.

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Pöml, Philipp, and Boris Burakov. "Study of the redistribution of U, Zr, Nb, Tc, Mo, Ru, Fe, Cr, and Ni between oxide and metallic phases in the matrix of a multiphase Chernobyl hot-particle extracted from a soil sample of the Western Plume." Radiochimica Acta 106, no. 12 (November 27, 2018): 985–90. http://dx.doi.org/10.1515/ract-2018-2957.

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Abstract A “hot particle” found 6 km west of the Chernobyl nuclear power plant 4 years after the Chernobyl severe nuclear accident was analysed by scanning electron microscopy and electron probe micro-analysis. The matrix of the particle consists of relics of partly molten UO2 nuclear fuel and two different phases of solidified U–Zr–O melt (U0.77Zr0.23O2 and U0.67Zr0.33O2). The particle also contains a unique metallic inclusion of a size of 30×22 μm. The inclusion is non-homogeneous and in some parts shows a dendrite-like structure. It consists of about 38 wt.% Fe, about 10 wt.% U, Mo, and Nb, about 5 wt.% Ru, Zr, Ni, and Cr, and small amounts of Tc (2 wt.%) and Si (0.4 wt.%). The presence of partly molten nuclear fuel suggests a local temperature exceeding 2850 °C. The metallic inclusion most likely formed when steel, fuel, and cladding reacted together and molten steel incorporated U, Zr, Nb, Tc, Mo, and Ru from molten fuel and cladding during a very fast high-temperature process. Fast quenching of the metallic and the oxide melt left no time for Tc and Mo to evaporate. Molten Zr was partly oxidised and acted as a buffer for O which caused the reduction of a fraction of the U. The data of this study support the previously reported supercritical nature of the Chernobyl explosion.
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SUDO, Ayako, Fumiki MIZUSAKO, Kuniyoshi HOSHINO, Takumi SATO, Yuji NAGAE, and Masaki KURATA. "Fundamental Study on Segregation Behavior in U–Zr–Fe–O System during Solidification Process." Transactions of the Atomic Energy Society of Japan 18, no. 3 (2019): 111–18. http://dx.doi.org/10.3327/taesj.j18.029.

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Nandan, Shambhavi, Florian Fichot, and Bruno Piar. "A simplified model for the quaternary U-Zr-Fe-O system in the miscibility gap." Nuclear Engineering and Design 364 (August 2020): 110608. http://dx.doi.org/10.1016/j.nucengdes.2020.110608.

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Khabensky, V. B., V. I. Almjashev, E. B. Shuvaeva, E. V. Krushinov, S. A. Vitol, A. A. Sulatsky, S. Yu Kotova, and V. V. Gusarov. "Experimental determination of spatial inversion pointof coexisting molten phases in the U-Zr-Fe-O system." Nuclear Propulsion Reactor Plants. Life Cycle Management Technologies., no. 3 (2021): 63–81. http://dx.doi.org/10.52069/2414-5726_2021_3_25_63.

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Dissertations / Theses on the topic "U-Zr-Fe-O"

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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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MAURIZI, ANNE. "Reactivite chimique a haute temperature dans le systeme (u, zr, fe, o). Contribution a l'etude de la zircone comme recuperateur de corium." Paris 6, 1996. http://www.theses.fr/1996PA066617.

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Notre travail s'est inscrit dans le cadre d'un programme de gestion des accidents nucleaires envisageant de placer sous la cuve des reacteurs un recuperateur destine a confiner le corium resultant de la fusion du coeur. D'apres l'etude bibliographique que nous avons realisee, la zircone stabilisee semble le materiau refractaire le mieux adapte pour remplir ce role, compte tenu des contraintes physico-chimiques, mecaniques et thermiques imposees au recuperateur. Nos recherches ont permis d'etablir la nature des interactions zircone/fer a haute temperature, et de determiner certaines donnees experimentales sur le quaternaire (u, zr, fe, o), systeme modele du corium. Dans un premier temps, nous avons etabli par spectrometrie de masse a haute temperature la position du liquidus pour une composition proche du corium dans le systeme (u, zr, o) a 2000c. La solubilite de l'oxygene dans un alliage (u, zr, o) liquide avec u/zr = 1,5 est de l'ordre de 7 atome %. En atmosphere oxydante, la reaction entre la zircone et le fer se traduit par la formation d'une solution solide zircone stabilisee - oxyde de fer. L'incorporation d'une quantite pouvant atteindre 10 atome % de fer stabilise la zircone cubique et modifie les parametres de maille. La valence et la localisation des atomes de fer dans la structure de type fluorine ont ete etudiees par spectroscopie mossbauer. La penetration du fer dans la zircone a ete mesuree en fonction du temps et de la temperature entre 1500 et 2400c apres chauffage par induction haute frequence, a la fois sur des creusets de laboratoire et des briques commerciales. Le processus est regi par une energie d'activation de l'ordre de 80 kj/mol. Les resultats demontrent que la zircone est capable d'absorber de facon efficace le fer oxyde.
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Brunel, Alan. "Propriétés thermodynamiques et thermophysiques des liquides à haute température : applications aux combustibles nucléaires." Electronic Thesis or Diss., Sorbonne université, 2022. http://www.theses.fr/2022SORUS426.

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Lors d’un accident grave impliquant la fusion du cœur d’un réacteur nucléaire à eau pressurisée, le combustible nucléaire va réagir avec la gaine en zircalloy qui l’enrobe et les matériaux de structure présents dans le cœur pour former un magma à haute température appelé corium. Suivant sa composition et sa température, le corium peut se stratifier dû à la présence d’un liquide métallique et d’un liquide oxyde non-miscibles. Selon la configuration de cette stratification, une concentration du flux de chaleur peut avoir lieu sur la paroi de la cuve, menaçant son intégrité et risquant un écoulement du corium hors de celle-ci. L’objectif de cette thèse est d’obtenir des données thermodynamiques et thermophysiques sur un corium prototypique, le système U-Zr-Fe-O. Les données thermodynamiques recueillies dans cette thèse sont liées à la définition de la lacune de miscibilité liquide et à la composition des liquides dans le système U-Zr-Fe-O et de ses sous-systèmes, en fonction de la composition et de la température. Des compositions d’intérêt sont sélectionnées suite à des calculs thermodynamiques réalisés par la méthode CALPHAD grâce à la base de données TAF-ID V13. Les échantillons relatifs à ces compositions ont subi des traitements thermiques et des analyses post-opératoires afin de mesurer les compositions des liquides et de les comparer aux calculs thermodynamiques. Une lacune de miscibilité liquide riche en fer et une autre riche en zirconium ont été mises en évidence dans le système Fe-Zr-O. Alors que les données obtenues sur la première lacune à 1990 °C et 2614 °C montrent un bon accord entre le calcul et l’expérience, les mesures sur la lacune riche en zirconium à 2420 °C et 2650 °C indiquent que le modèle sous-estime la quantité de zirconium dans le liquide métallique et, à l’inverse, la surestime dans le liquide oxyde. Les études réalisées sur le système UO2-Zr-Fe à 2423 °C montrent que la présence de la lacune de miscibilité liquide et la composition des liquides dépendent grandement de la quantité de fer dans le système, du rapport U/Zr et du degré d’oxydation du corium. De plus, le modèle tend à sous-estimer la fraction molaire de zirconium dans le liquide métallique au profit du fer, et à la surestimer dans le liquide oxyde. Enfin, le modèle sous-estime grandement la solubilité de l’oxygène dans le liquide métallique. L’obtention de données thermophysiques a pu être réalisée grâce à l’amélioration du banc expérimental ATTILHA, rendant possible l’étude de liquides sensibles à l’oxygène ou radioactifs à hautes températures via un chauffage laser. Ce banc a permis de mesurer des valeurs expérimentales de température de liquidus et de transition eutectique sur le système Zr-O dans le domaine riche en oxygène. De plus, le développement de la lévitation aérodynamique sur ce banc permit l’étude de la masse volumique de liquides Zr-Fe2O3 et Zr-UO2 entre 1884 °C et 2268 °C pour différentes fractions molaires de zirconium. Les résultats de masse volumique des liquides Zr-Fe2O3 ont permis d’affiner des mesures de tension de surface réalisées sur le banc VITI-MBP au CEA Cadarache. Ces mesures confirmèrent les propriétés surfactantes de l’oxygène sur ces liquides. Les données expérimentales recueillies durant cette thèse pourront servir à alimenter les codes de calcul afin de mieux prédire le comportement du corium et le déroulement des accidents graves
During a severe accident involving the meltdown of the core of a pressurized water nuclear reactor, the nuclear fuel will react with the zircalloy cladding around it and the structural materials of the core to make a high temperature magma called corium. Depending on its composition and its temperature, the corium can stratify because of two non-miscible metallic and oxidic liquids. For some stratification configurations, the heat flow can focus on the vessel’s wall, threatening its integrity with a corium flowing outside of it. The aim of this thesis is to collect thermodynamic and thermophysic data on a prototypical corium, the U-Zr-Fe-O system. The thermodynamic data collected in this thesis are related to the definition of the liquid miscibility gap and the compositions of the liquids in the U-Zr-Fe-O system and its sub-systems, depending on the composition and the temperature. Compositions of interest were selected after performing thermodynamic calculation by the CALPHAD method with the TAF-ID V13 database. The corresponding samples underwent heat treatments and post-treatment analyses to measure the compositions of the liquids and to compare them to thermodynamic calculations. An iron rich liquid miscibility gap and a zirconium rich one were highlighted in the Fe-Zr-O system. Although calculations were in agreement with data from the first miscibility gap at 1990 °C, measurements in the zirconium rich miscibility gap at 2420 °C and 2650 °C reveal an underestimation of the zirconium quantity in the metallic liquid and its overestimation in the oxidic liquid by the model. Studies on the UO2-Zr-Fe system at 2423 °C show that the liquid miscibility gap definition and the compositions of the liquids depend on the quantity of iron in the system, the U/Zr ratio and corium oxidation degree. Furthermore, the zirconium molar fraction is underestimated by the model in the metallic liquid to the benefit of iron, and is overestimated in the oxidic liquid. Finally, the oxygen solubility in the metallic liquid is underestimated by the model. Thermophysic data were collected thanks to the improvement of the ATTILHA experimental setup, allowing the study of oxygen sensitive or radioactive liquids at high temperature by using a laser heating. Experimental values on liquidus and eutectic transformation temperatures of the oxygen-rich domain of the Zr-O system were acquired with this setup. Furthermore, the development of the aerodynamic levitation allows us the investigation liquids’densities for the Zr-Fe2O3 and the Zr-UO2 systems between 1884 °C and 2268 °C for different zirconium molar fractions. Densities of liquids from the Zr-Fe2O3 system were used to refine surface tension values acquired on the VITI-MBP setup at CEA Cadarache. These values confirmed the surfacting properties of the oxygen on these liquids. The experimental data collected during this thesis will be used to feed the databases and to improve the forecast of the corium’s behavior during a severe accident
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Book chapters on the topic "U-Zr-Fe-O"

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McFarland, Ben. "Unfolding the Periodic Table." In A World From Dust. Oxford University Press, 2016. http://dx.doi.org/10.1093/oso/9780190275013.003.0007.

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Our starting point is not hidden, nor is it far off. It is not an extreme place like Mono Lake or Freswick Castle, but it is a central concept expressed on a single page. The periodic table is the center of chemistry, and therefore of this book. You can spot it at a distance from its vaguely cathedral-like shape. You can see the chemical symbols that it contains on magnets and T-shirts and restaurant signs. Its regular columns are not quite symmetric, but that is because it has been twisted out of its natural shape by the contingencies of history. Rearrange it just a little and a simple mathematical pattern appears. To see this pattern, imagine that the periodic table is made out of beads on an abacus, arranged in the familiar U shape. Then push all the beads to the left: … Row 1 = H- He Row 2 = Li- Be- B- C- N- O- F- Ne Row 3 = Na- Mg- Al- Si- P- S- Cl- Ar Row 4 = K- Ca- Sc- Ti- V- Cr- Mn- Fe- Co- Ni- Cu- Zn- Ga- Ge- As- Se- Br- Kr Row 5 = Rb- Sr- Y- Zr- Nb- Mo- Tc- Ru- Rh- Pd- Ag- Cd- In- Sn- Sb- Te- I- Xe … By row, there are 2, 8, 8, 18, and 18 elements. The pattern continues in the rows below, but it is obscured by the fact that on most tables 14 elements have been moved out of the sixth and seventh rows. On the table here I have put them where they belong. These rows have 32 elements each. This can be simplified even more. The rows increase, first by 2, then by 6 more (2 + 6 = 8), then by 10 more (2 + 6 + 10 = 18), then by 14 (2 + 6 + 10 + 18 = 32). The series 2, 6, 10, 14 is the doubles of counting up by odd numbers: 1, 3, 5, 7. Put another way, each row is equal to 2n + 1 with n = integers from 0.
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Conference papers on the topic "U-Zr-Fe-O"

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Carénini, L., and F. Fichot. "The Impact of Transient Behavior of Corium in the Lower Head of a Reactor Vessel for In-Vessel Melt Retention Strategies." In 2016 24th International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2016. http://dx.doi.org/10.1115/icone24-60598.

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One of the main goals of severe accident management strategies is to mitigate radiological releases to people and environment. To choose the most appropriate strategy, one needs to know the probability of its success taking into account the associated uncertainties. In the field of corium and debris behavior and coolability, research programs are still on going and the possibilities to efficiently cool and retain corium and debris inside the Reactor Pressure Vessel (RPV) then inside the containment are difficult to evaluate. This leads to uncertainties in safety assessments particularly when margins to RPV or containment failure are too weak. In Vessel Melt Retention (IVMR) strategies for Light Water Reactors (PWR, BWR, VVER) intend to stabilize and retain the core melt in the RPV (as it happened during the TMI-2 accident). This would reduce significantly the threats to the last barrier (the containment) and therefore reduce the risk of release of radioactive elements to the environment. This type of Severe Accident Management (SAM) strategy has already been incorporated recently in the SAM guidance (SAMG) of several operating medium size Light Water Reactors (reactor below 500MWe (like VVER440)) and is part of the SAMG strategies for some Gen III+ PWRs of higher power like the AP1000. A European project coordinated by IRSN and gathering 23 organizations (Utilities, Technical Support Organizations, Nuclear Power Plant vendors, Research Institutes…) has been launched in 2015 with as main objective the evaluation of feasibility of IVMR strategies for Light Water Reactors (PWR, VVER, BWR) of total power around 1000MWe (which represent a significant part of the European Nuclear Power Plants fleet). This paper intends to show how it is possible to introduce transient evolutions of the stratified corium pool in the evaluation of the heat flux profile along the vessel wall. Indeed, due to chemical reactions in the U–Zr–O–Fe molten pool, separation between non-miscible metallic and oxide phases may occur, modifying the thermal load applied to the RPV. If stabilized stratified corium configurations are well defined and modeled, transient evolutions of material layers in the corium pool are still difficult to predict. The evaluations presented are based on calculations performed with the severe accident integral code ASTEC (Accident Source Term Evaluation Code) for a typical PWR reactor. The modeling of transient evolution of corium layers leads to configurations with a thin light metal layer on top of the oxidic one, increasing the so called “focusing effect” (intense heat fluxes on the RPV walls adjacent to the top metal layer). A sensitivity study on some uncertain parameters is proposed to evaluate the impact on the kinetics of layers inversion. Depending on the duration of these transient heat fluxes, the mechanical strength of the RPV could be challenged.
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