Relevant bibliographies by topics / Corium

Contents

  1. Journal articles
  2. Dissertations / Theses
  3. Books
  4. Book chapters
  5. Conference papers
  6. Reports

Academic literature on the topic 'Corium'

Author: Grafiati
Published: 4 June 2021
Last updated: 26 July 2025

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Journal articles on the topic "Corium"

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Tursinah, Rasito, Marisa Variastuti, Rakotovao Lovanantenaina Omega, Asril Pramutadi Andi Mustari, and Sidik Permana. "MPS SIMULATION ON THE CORIUM MELT FLOW IN CASE OF REACTOR ACCIDENT." GANENDRA Majalah IPTEK Nuklir 26, no. 2 (2024): 91. http://dx.doi.org/10.55981/gnd.2023.6829.

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A simulation model has been made for molten corium in a nuclear reactor using the Moving Particle Semi-Implicit (MPS) method. By setting the value of dynamic viscosity and temperature of corium, simulations are carried out to display the pressure profile and flow velocity of the corium fluid that falls from the RPV to the plenum. In the first simulation to observe the pressure and velocity profile of the corium in the plenum, three conditions were made: the plenum was empty; the plenum was filled with corium fluid, and the plenum was filled with debris. The second simulation was carried out to
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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 (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
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Baklanova, Yu Yu, O. S. Bukina O. S. Bukina O. S. Bukina, and V. V. Baklanov. "METHODOLOGY FOR THE STUDY OF CORIUM AGING PROCESSES." NNC RK Bulletin, no. 1 (April 1, 2025): 104–12. https://doi.org/10.52676/1729-7885-2025-1-104-112.

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To date, the corium research is one of the main issues in the framework of improving nuclear safety and is one of the tasks of conducting a successful procedure to eliminate the consequences of an accident with a core meltdown at the NPP. One of the important tasks for the procedure of eliminating the consequences of an accident at the NPP is to understand the physical state of the core melt of an emergency reactor (corium) in order to make decisions on its removal from the contents and further handling. The difficulty in assessing the structure and properties of the corium, which undergo the
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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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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 th
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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) (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 in
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Skakov, Mazhyn, Viktor Baklanov, Assan Akaev, et al. "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 (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
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Journeau, Christophe, Laurence Aufore, Léonie Berge, et al. "Corium-Sodium and Corium-Water Fuel-Coolant-Interaction Experimental Programs for the PLINIUS2 Prototypic Corium Platform." Nuclear Technology 205, no. 1-2 (2018): 239–47. http://dx.doi.org/10.1080/00295450.2018.1479580.

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Skakov, Mazhyn, Viktor Baklanov, Maxat Bekmuldin, et al. "Results of experimental simulation of interaction between corium of a nuclear reactor and sacrificial material (Al<sub>2</sub>O<sub>3</sub>) 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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&lt;abstract&gt; &lt;p&gt;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
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Bukina, O. S., A. D. Grechanik, E. A. Kozhakhmetov, I. M. Kukushkin, and Yu Yu Baklanova. "INVESTIGATION OF URANIUM AND ZIRCONIUM BASED SOLID SOLUTIONS." NNC RK Bulletin, no. 4 (December 30, 2020): 69–76. http://dx.doi.org/10.52676/1729-7885-2020-4-69-76.

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In the IAE Branch RSE NNC RK at the VCG-135 test-bench within the framework of the commercial projects and scientific and technical program entitled “Study of the corium prototype properties of various compositions” small-scale experiments are carried out to obtain corium prototypes of various compositions. Physical and mechanical properties, phase and elemental composition of corium prototype samples resulted from high-temperature experiments on test-benches are being investigated based on the Material Testing Department.The work was aimed at identifying solid solutions based on uranium and z
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Dissertations / Theses on the topic "Corium"

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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 s
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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 obser
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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. C
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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&gt;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 conditi
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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&gt; 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
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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
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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 favo
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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 pa
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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 p
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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'instab
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Books on the topic "Corium"

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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. 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. Division of Safety Issue Resolution, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, 1991.

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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. Division of Safety Issue Resolution, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, 1991.

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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. 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. Division of Systems Research, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, 1994.

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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. 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. 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. 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. Farming Press Books, 1989.

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Moorcock, Michael. Corum. Millennium, 1992.

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Book chapters on the topic "Corium"

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Bährle-Rapp, Marina. "Corium." In Springer Lexikon Kosmetik und Körperpflege. 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, et al. "Corium." In Encyclopedia of Entomology. 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. 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. 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. 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. Springer Japan, 2010. http://dx.doi.org/10.1007/978-4-431-99779-5_43.

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Siddiqui, Osman, Abdallah Balabaid, Osamah Al-Gazlan, and Afaque Shams. "Numerical Prediction of Flow and Heat Transfer in a Molten Corium Pool." In Lecture Notes in Mechanical Engineering. Springer Nature Switzerland, 2024. http://dx.doi.org/10.1007/978-3-031-64362-0_46.

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Khurshid, Ilyas, Imran Afgan, and Yacine Addad. "A CFD Approach to Mimic the Molten Corium-Concrete Interaction Phenomena: Effects of the Thermal Boundary Conditions." In Lecture Notes in Mechanical Engineering. Springer Nature Switzerland, 2024. http://dx.doi.org/10.1007/978-3-031-64362-0_35.

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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. 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. Springer Berlin Heidelberg, 1992. http://dx.doi.org/10.1007/978-3-642-77401-0_53.

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Conference papers on the topic "Corium"

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Pegarkov, Alex, Shawn Somers-Neal, Abubaker Alatrash, Edgar Matida, Vinh Tang, and Tarik Kaya. "Numerical Investigation of Solidification of Corium in an Initially Emptied Vertical Pipe." In Mathematics and Computation 2021. American Nuclear Society, 2021. https://doi.org/10.13182/xyz-33648.

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Journeau, C., and J. M. Seiler. "Interaction Corium-Béton." In Premières conséquences du REX de Fukushima sur l’exploitation des réacteurs et installations nucléaires. 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. 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, depend
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Xu, Yihua, Ryo Yokoyama, and Shunichi Suzuki. "Experimental Study on Metal Jet Spreading on Substrate With Roughness." In 2024 31st International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2024. http://dx.doi.org/10.1115/icone31-135231.

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Abstract The behavior of molten corium, particularly when it impinges upon the containment substrate and the further spreading, is important in determining the integrity of the containment structures. In nuclear power plants (NPPs), substrates typically display a variety of surface textures; however, the impact of substrate roughness on corium behavior remains inadequately explored. In this study, the influence of substrate roughness on the corium spreading behavior is investigated. A series of scaled experiments were conducted using Wood’s metal as corium simulant impacting substrates of diff
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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
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Journeau, Christophe, Viviane Bouyer, Nathalie Cassiaut-Louis, 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 we
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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 perfo
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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 c
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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 t
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Reports on the topic "Corium"

1

Huhtiniemi, I., H. Hohmann, and D. Magallon. FCI experiments in the corium/water system. Office of Scientific and Technical Information (OSTI), 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), 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), 1987. http://dx.doi.org/10.2172/6594223.

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Louie, David, Yifeng Wang, Rekha R. Rao, et al. A New Method to Contain Molten Corium in Catastrophic Nuclear Reactor Accidents. Office of Scientific and Technical Information (OSTI), 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), 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), 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), 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), 1988. http://dx.doi.org/10.2172/7013321.

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Winston, Philip. Management of the Three Mile Island, Unit 2, Accident Corium and Severely Damaged Fuel Debris. Office of Scientific and Technical Information (OSTI), 2025. https://doi.org/10.2172/2564194.

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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), 1995. http://dx.doi.org/10.2172/115071.

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