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Journal articles on the topic 'Corium'

Author: Grafiati

Published: 4 June 2021

Last updated: 8 June 2024

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1

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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2

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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3

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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4

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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5

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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6

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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7

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 (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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<abstract> <p>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.</p> </abstract>

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8

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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9

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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10

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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11

Smirnov, Anton D., Ekaterina V. Bogdanova, Pavel A. Pugachev, Ivan S. Saldikov, Mikhail Yu Ternovykh, Georgy V. Tikhomirov, Hiroki Takezawa, Takeshi Muramoto, Jun Nishiyama, and Toru Obara. "Neutronic modeling of a subcritical system with corium particles and water (from international benchmark)." Nuclear Energy and Technology 6, no. 3 (September 16, 2020): 155–60. http://dx.doi.org/10.3897/nucet.6.57742.

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After the accident at the Fukushima Daiichi NPP, the attention of the scientific community is riveted on how the consequences are being eliminated. Removing corium – a lava-like resolidified mixture of nuclear fuel with other structural elements of the reactor – remains the most difficult task, the solution of which can take several decades. It is extremely important to exclude the occurrence of any emergency processes during the removal of corium. The purpose of this work was to solve a coordinated hydrodynamic and neutronic problem characterized by a large number of randomly oriented and irregularly located corium particles in water as part of the development of a benchmark for this class of problems. Monte Carlo-based precision codes were used to perform a neutronic analysis. The positions of corium particles were determined from the numerical simulation results. The analysis results obtained using the codes involved showed good agreement for all the states considered. It was shown that the modern neutronic codes based on the Monte Carlo method successfully cope with the geometric formation and solution of the problem with a nontrivial distribution of corium particles in water. The results of the study can be used to justify the safety of corium handling procedures, including its extraction from a damaged power unit.

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12

Ryzhov, S. N., E. V. Bogdanova, A. A. Ryzhkov, P. A. Pugachev, G. V. Tikhomirov, M. Yu Ternovykh, and T. B. Aleeva. "Analysis of Methods and Technologies for Composition Assessing of Corium Formed as a Result of the Fukushima Daiichi NPP Accident." Global Nuclear Safety, no. 3 (August 31, 2022): 5–21. http://dx.doi.org/10.26583/gns-2022-03-01.

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This paper analyzes the methods and technologies for assessing the method of formation, composition, characteristics and features of corium, which is a mixture of nuclear and structural materials of the nuclear reactor core, formed as a result of an accident accompanied by partial or complete core melting. The study is based on data from the study of corium formed as a result of the accident at the Fukushima Daiichi nuclear power plant, which are in the public domain and are the result of the work of many scientific organizations around the world. Corium research is one of the main issues in the framework of improving nuclear safety in the future and is one of the objectives of the successful procedure for eliminating the consequences of the accident at the Fukushima Daiichi nuclear power plant. Without a detailed analysis of the neutronic, materials science, gravimetric and other characteristics of the corium, as well as the creation of a complex model of the corium that combines these data, it is impossible to organize an efficient and safe process for removing nuclear materials from the damaged units of the Fukushima Daiichi nuclear power plant. The objective of this work is to combine the existing research results into a data set that allows modeling of the corium using neutronic calculation codes and includes such data as the size, density and morphology of corium samples and their approximate nuclide composition. Such modeling allows not only to perform tasks related to increasing the level of safety in the implementation of the procedure for eliminating the consequences of the accident at the Fukushima Daiichi nuclear power plant, but also to serve as an international benchmark for modeling a mixture containing nuclear materials.

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13

Baklanov, V. V., A. V. Gradoboev, and V. S. Zhdanov. "Development of the Technique to Simulate Residual Heading Corium Prototype." Applied Mechanics and Materials 770 (June 2015): 130–36. http://dx.doi.org/10.4028/www.scientific.net/amm.770.130.

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The paper addresses simulating technique of residual energy release inside of corium prototype based on the usage of plasmotron; describes plasmotron design and its manufacturing features; considers research results of plasmotron interaction with the corium prototype.

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14

Min, B. T., S. W. Hong, J. H. Kim, I. K. Park, and H. D. Kim. "Dominant Factor for the Occurrence of a Steam Explosion." Defect and Diffusion Forum 273-276 (February 2008): 388–93. http://dx.doi.org/10.4028/www.scientific.net/ddf.273-276.388.

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For the study of a steam explosion phenomenon in a nuclear reactor, prototypic corium, a mixture of UO2 and ZrO2 was melted in a cold crucible by applying an induction heating technique. The molten corium was then poured into cold water. It was fragmented into very small particles, so called debris, which enables a very rapid heat transfer to the water. Some cases led to steam explosions by thermal expansion of the water. After the tests, all the debris particles were dried and classified by their size. From the analysis by using EPMA, it was shown that the particles generated by a steam explosion had fine and irregular forms. It is known that real corium (including UO2) hardly leads to a steam explosion, different from pure ZrO2 or metal. A reason for this was previously suggested in that the corium generated hydrogen gas during melt-water interaction, and it enclosed the melt drops to prevent a direct contact of the corium and water. In order to confirm this fact, the debris particles were analyzed with ICP-AES for their typical element contents, EPMA for the homogeneity of the solid solution, XRD for the chemical compounds, and TGA and hydrogen reduction analysis for the percentage of the debris oxidation and reduction. These analyses showed that hydrogen was not directly related to steam explosion. Meanwhile, the material characteristics of the corium compositions are newly suggested to be the most probable reason for the occurrence of a steam explosion so far.

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15

Løken, S. B., I. Skrede, and T. Schumacher. "The Helvella corium species aggregate in Nordic countries – phylogeny and species delimitation." Fungal Systematics and Evolution 5, no. 1 (June 1, 2020): 169–86. http://dx.doi.org/10.3114/fuse.2020.05.11.

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Mycologists have always been curious about the elaborate morphotypes and shapes of species of the genus Helvella. The small, black, cupulate Helvella specimens have mostly been assigned to Helvella corium, a broadly defined morpho-species. Recent phylogenetic analyses, however, have revealed an aggregate of species hidden under this name. We performed a multispecies coalescent analysis to re-assess species limits and evolutionary relationships of the Helvella corium species aggregate in the Nordic countries. To achieve this, we used morphology and phylogenetic evidence from five loci – heat shock protein 90 (hsp), translation elongation factor 1-alpha (tef), RNA polymerase II (rpb2), and the 5.8S and large subunit (LSU) of the nuclear ribosomal DNA. All specimens under the name Helvella corium in the larger university fungaria of Norway, Sweden and Denmark were examined and barcoded, using partial hsp and/or rpb2 as the preferential secondary barcodes in Helvella. Additional fresh specimens were collected in three years (2015–2018) to obtain in vivo morphological data to aid in species discrimination. The H. corium species aggregate consists of seven phylogenetically distinct species, nested in three divergent lineages, i.e. H. corium, H. alpina and H. pseudoalpina sp. nov. in the /alpina-corium lineage, H. alpestris, H. macrosperma and H. nannfeldtii in the /alpestris-nannfeldtii lineage, and H. alpicola as a weakly supported sister to the /alpestris-nannfeldtii lineage. Among the seven species, the ribosomal loci expressed substantial variation in evolutionary rates, suggesting care in the use of these regions alone in delimitation of Helvella species. Altogether, 469 out of 496 available fungarium specimens were successfully barcoded.

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FUJIMARU, Atsushi, and Atsushi TANIGUCHI. "Development of Corium shield." Proceedings of the National Symposium on Power and Energy Systems 2018.23 (2018): A113. http://dx.doi.org/10.1299/jsmepes.2018.23.a113.

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17

Veshchunov, M. S., K. Mueller, and A. V. Berdyshev. "Molten corium oxidation model." Nuclear Engineering and Design 235, no. 22 (November 2005): 2431–50. http://dx.doi.org/10.1016/j.nucengdes.2005.05.003.

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18

Lomperski, S., and M. T. Farmer. "Corium crust strength measurements." Nuclear Engineering and Design 239, no. 11 (November 2009): 2551–61. http://dx.doi.org/10.1016/j.nucengdes.2009.06.013.

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19

Abalin, S. S., V. G. Asmolov, V. D. Daragan, E. K. D’yakov, A. V. Merzlyakov, and V. Yu Vishnevsky. "Corium kinematic viscosity measurement." Nuclear Engineering and Design 200, no. 1-2 (August 2000): 107–15. http://dx.doi.org/10.1016/s0029-5493(00)00238-7.

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20

Cognet, G., H. Alsmeyer, W. Tromm, D. Magallon, R. Wittmaack, B. R. Sehgal, W. Widmann, et al. "Corium spreading and coolability." Nuclear Engineering and Design 209, no. 1-3 (November 2001): 127–38. http://dx.doi.org/10.1016/s0029-5493(01)00395-8.

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21

Zubekhina, Bella, Boris Burakov, Ekaterina Silanteva, Yuri Petrov, Vasiliy Yapaskurt, and Dmitry Danilovich. "Long-Term Aging of Chernobyl Fuel Debris: Corium and “Lava”." Sustainability 13, no. 3 (January 21, 2021): 1073. http://dx.doi.org/10.3390/su13031073.

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Samples of Chernobyl fuel debris, including massive corium and “lava” were collected inside the Chernobyl “Sarcophagus” or “Shelter” in 1990, transported to Leningrad (St. Petersburg) and stored under laboratory conditions for many years. In 2011 aged samples were visually re-examined and it was confirmed that most of them remained intact, although some evidence of self-destruction and chemical alteration were clearly observed. Selected samples of corium and “lava” were affected by static leaching at temperatures of 25, 90 and 150 °C in distilled water. A normalized Pu mass loss (NLPu) from corium samples after 140 days was noted to be 0.5 g/m2 at 25 °C and 1.1 g/m2 at 90 °C. For “lava” samples NLPu was 2.2–2.3 g/m2 at 90 °C for 140 days. The formation of secondary uranyl phases on the surface of corium and “lava” samples altered at 150 °C was confirmed. The results obtained are considered as an important basis for the simulation of fuel debris aging at Fukushima Daiichi nuclear power plant (NPP).

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Ławrynowicz, Maria, and Andrzej Radwański. "A contribution to the morphology and ecology of Mycenastrum corium (Agaricales)." Acta Mycologica 41, no. 1 (December 23, 2013): 73–78. http://dx.doi.org/10.5586/am.2006.011.

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An interesting collection of <em>Mycenastrum corium</em> from Suwałki Region (NE Poland) close to the Russian and Lithuenian frontiers is presented in this paper. Two specimens were found ca. 20 cm under the soil surface. Macro- and micromorphological features are compared with those of <em>Mycenastrum corium</em> growing at the surface.

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23

Skakov, Mazhyn K., Nurzhan Ye Mukhamedov, Alexander D. Vurim, and Ilya I. Deryavko. "Temperature Dependence of Thermophysical Properties of Full-Scale Corium of Fast Energy Reactor." Science and Technology of Nuclear Installations 2017 (2017): 1–7. http://dx.doi.org/10.1155/2017/8294653.

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For the first time the paper determines thermophysical properties (specific heat capacity, thermal diffusivity, and heat conductivity) of the full-scale corium of the fast energy nuclear reactor within the temperature range from ~30°С to ~400°С. Obtained data are to be used in temperature fields calculations during modeling the processes of corium melt retention inside of the fast reactor vessel.

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24

Eichler, Wolfgang, Christine Eisenbeiss, Jan Schumacher, Stefan Klaus, Rolf Vogel, and Karl Friedrich Klotz. "Changes of interstitial fluid volume in superficial tissues detected by a miniature ultrasound device." Journal of Applied Physiology 89, no. 1 (July 1, 2000): 359–63. http://dx.doi.org/10.1152/jappl.2000.89.1.359.

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We evaluated the changes of tissue layer thickness in circumscribed superficial tissue areas with a 10-MHz A-mode and a 20-MHz B-mode ultrasound device under alterations in body posture and plasma volume to detect fluid shifts between the different compartments. In 20 male volunteers, we measured tissue thickness by A mode and corium and subcutis thickness by B mode at the forehead before and 30 min after three procedures: change from upright to supine position (P1); change from upright to 30° head-down-tilt position (P2); infusion of 10 ml/kg body wt of Ringer solution (P3). We found a significant correlation between baseline tissue thickness and the sum of corium and subcutis thicknesses ( r = 0.75, P < 0.01). The changes of body posture and plasma volume resulted in significant increases of tissue thickness (P1, 2.9%; P2, 11.6%; P3, 5.8%) and corium thickness (P1, 4.7%; P2, 8.1%; P3, 9.1%) but not of the sum of chorium and subcutis thicknesses. We conclude that fluid shifts from the intravascular to the extravascular compartment are detectible by evaluating corium thickness with a B-mode, or more easily tissue thickness with an A-mode, ultrasound device.

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Yokoyama, Ryo, Shunichi Suzuki, Koji Okamoto, and Masaru Harada. "Scale effect of amount of molten corium and outlet diameters on corium spreading." Progress in Nuclear Energy 130 (December 2020): 103535. http://dx.doi.org/10.1016/j.pnucene.2020.103535.

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Zubekhina, Bella Yu, Boris E. Burakov, Oksana G. Bogdanova, and Yuriy Yu Petrov. "Leaching of 137Cs from Chernobyl fuel debris: corium and “lava”." Radiochimica Acta 107, no. 12 (November 26, 2019): 1155–60. http://dx.doi.org/10.1515/ract-2019-0009.

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Abstract Samples of Chernobyl fuel debris such as corium and “lava” had been studied using repeated static leach test MCC-1 at temperature of 25 and 90 °C in distilled water and simulated seawater. A normalized 137Cs mass loss (NLCs) estimated for corium samples after 168 days in distilled and seawater was 3.2–3.5 g/m2 at 25 °C and 113–114 g/m2 at 90 °C. For “lava” samples NLCs varied from 1.4 to 13.2 g/m2 at 90 °C for 56 days (in distilled and seawater) and from 0.1 to 0.4 g/m2 at 25 °C in seawater for 140 days. Chemical durability of Chernobyl “lava” in distilled and seawater evaluated using 137Cs specific activity in leachates is higher than one for corium. Further study is proposed in order to obtain more quantitative data.

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27

Seiler, J. M., and J. Ganzhorn. "Viscosities of corium–concrete mixtures." Nuclear Engineering and Design 178, no. 3 (December 1997): 259–68. http://dx.doi.org/10.1016/s0029-5493(97)00232-x.

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28

Hardie, Susie M. L., Ian G. McKinley, Steve Lomperski, Hideki Kawamura, and Tara M. Beattie. "Management options for Fukushima corium." Progress in Nuclear Energy 92 (September 2016): 260–66. http://dx.doi.org/10.1016/j.pnucene.2015.07.017.

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29

Sulatsky, A. A., S. A. Smirnov, V. S. Granovsky, V. B. Khabensky, E. V. Krushinov, S. A. Vitol, S. Yu Kotova, et al. "Oxidation kinetics of corium pool." Nuclear Engineering and Design 262 (September 2013): 168–79. http://dx.doi.org/10.1016/j.nucengdes.2013.04.025.

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30

Ramacciotti, Muriel, Christophe Journeau, François Sudreau, and Gérard Cognet. "Viscosity models for corium melts." Nuclear Engineering and Design 204, no. 1-3 (February 2001): 377–89. http://dx.doi.org/10.1016/s0029-5493(00)00328-9.

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31

Toleubekov, K. O., A. S. Khazhidinov, and A. S. Akaev. "MODELING OF THE INDUCTION HEATING FOR IMITATION DECAY HEAT IN THE CORIUM DURING THE INTERACTION WITH HEAT-RESISTANT MATERIALS." NNC RK Bulletin, no. 1 (May 1, 2021): 9–14. http://dx.doi.org/10.52676/1729-7885-2021-1-9-14.

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This work is devoted to modeling of the induction heating the corium melts pouring on in the trap. The results nonstationary thermophysical calculation of the temperature field of the corium and refractory blocks of the melt trap are presented in the article. In the process of work, 2D model of the selected the melt trap area was created in the program ANSYS and the thermophysical model was validated by comparison the calculated and experimental data of the experiment.

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Maurin, L., P. Ferdinand, V. Bouyer, A. Denoix, G. Jouvin, S. Rougeault, C. Journeau, D. Molina, P. Tena, and Y. Ouerdane. "Remote monitoring of Molten Core-Concrete Interaction experiment with Optical Fibre Sensors & perspectives to improve nuclear safety – DISCOMS project." EPJ Web of Conferences 225 (2020): 08004. http://dx.doi.org/10.1051/epjconf/202022508004.

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The DISCOMS project (Distributed Sensing for Corium Monitoring and Safety) aimed at providing innovative solutions not requiring local electrical power supplies, for remote monitoring of a severe nuclear accident. The solutions are based on both long length SPNDs (Self Powered Neutron Detectors) and on distributed OFSs (Optical Fibre Sensors) capable to detect the onset of a severe accident, the corium pouring on the containment building concrete basemat, and its interaction with the concrete floor under the reactor vessel, until it spreads in the core catcher (EPR case). This paper mainly focuses on these last three detection targets achievable with distributed OFSs. It is based on the results of a Molten Core & Concrete Interaction (MCCI) experiment, namely VULCANO, held in June 2018 with a concrete crucible equipped with overall ~ 180 m long optical fibre sensing cables. This small scale experiment (50 kg of prototypical corium) has demonstrated the ability of distributed OFSs to remotely provide useful data during the MCCI run: i) temperature profiles images up to about 580°C (single wavelength Raman DTS reflectometer) until cooling down to room temperature, ii) high spatial-resolution frequency shifts profiles, due to combined (non-selective) strain and temperature influences (Rayleigh OFDR and Brillouin reflectometers), and iii) cables lengths ablated by the corium on sections weakened by the temperature (Raman DTS, Rayleigh OFDR, telecom and photon counting reflectometers).

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García-Lascuráin, Alma A., Gabriela Aranda-Contreras, Margarita Gomez-Chavarin, Ricardo Gómez, Adriana Méndez-Bernal, Gabriel Gutiérrez-Ospina, and María Masri. "Tratamiento de la laminitis crónica en equinos utilizando células troncales mesenquimales alogénicas de la médula ósea." Revista Mexicana de Ciencias Pecuarias 12, no. 3 (December 15, 2021): 721–41. http://dx.doi.org/10.22319/rmcp.v12i3.5765.

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Chronic laminitis is a disabling condition that affects the laminar corium of the horse’s hooves. Commonly, it develops as a collateral injury of numerous primary systemic diseases. It is believed that the critical physiopathological event that renders a hoof laminitic is the loss of mesenchymal stem cells. This loss greatly impairs the ability of the laminar corium to regenerate. Although previous work provides credibility to this notion, there remain unsettled issues that must be addressed before accepting it as a well-founded fact. Here, it was reexamined the central tenet of the physiopathological model of laminitis by infusing allogeneic bone marrow-derived mesenchymal stem cells (ABM-MSCs), through the digital palmar vein, into the hooves of horses afflicted by chronic laminitis. Horses were clinically monitored during 6 mo by evaluating them monthly using the lameness-modified Obel-Glasgow’s scale and hooves thermography. Venograms and lamellar biopsies were taken at the beginning and at the end of the study period to gathered evidence on vascular remodeling and laminar corium regeneration. The results showed that ABM-MSCs infusion promotes vascular remodeling and laminar corium regeneration, further supporting that the loss of stem cells is the critical event leading to chronic laminitis. This work also demonstrated that the infusion of ABM-MSCs is safe since the treated horses did not develop local or systemic, negative clinical manifestations attuned with rejection reactions, at least during the 6-mo period they were follow up and under the therapeutic scheme proposed.

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34

Tisseur, D., M. Cavaro, F. Rey, K. Paumel, N. Chikhi, J. Delacroix, P. Fouquart, R. Le Tellier, and V. Bouyer. "Study of online measurements techniques of metallic phase spatial distribution into a corium pool." EPJ Web of Conferences 225 (2020): 08003. http://dx.doi.org/10.1051/epjconf/202022508003.

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In the context of in-vessel retention (IVR) strategy in order to better assess the risk of reactor vessel failure, the knowledge related to the kinetics of immiscible liquid phases stratification phenomenon needs to be further improved. So far, only one medium-scale experiment (MASCA-RCW, in the frame of the OECD MASCA program) gives direct information regarding the transient relocation of metal below the oxide phase through post-mortem measurements. No experimental characterization of the stratification inversion kinetics when heavy metal becomes lighter and relocates at the top exists. Further investigation of these hydrodynamic and thermochemical processes could be made possible thanks to on line instrumentation enabling to follow displacement of oxidic and metallic phases into the corium pool. At CEA Cadarache, studies are under progress to set up innovative technologies for corium stratification monitoring which would be integrated to a cold crucible induction melting furnace. Based on space and time resolution specifications, three on-line measurements techniques were selected and studied. The first one is an ultrasonic technique using a refractory material waveguide and based on a time-of-flight measurement. We present the feasibility approach with the preliminary results obtained during experiments at high temperature on VITI facility. The second method consists in electromagnetic characterization of the corium pool thanks to an excitation by a magnetic field induced by surroundings coils and measurement of magnetic response by sensors placed around the crucible. A modelling study has enabled to define an appropriate experimental configuration. An experimental set up has also been tested to verify the calculation results. The third technique is 2D X-rays imaging. A feasibility study for a real-time X-ray imagingwith a framerate of 1 image/s has been performed using home-made simulation software MODHERATO, accounting forscattering, based on corium behavior previsions. Results on thedetection of interfaces between different type of corium phases(oxide, light metal, heavy metal) are shown.

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Beshta, S. V., V. S. Granovski, and A. A. Sulatski. "Water Boiling on Corium Melt Surface." Heat Transfer Research 30, no. 7-8 (1999): 515–21. http://dx.doi.org/10.1615/heattransres.v30.i7-8.120.

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Cognet, G., J. M. Seiler, I. Szabo, J. C. Latche, B. Spindler, and J. M. Humbert. "La récupération du corium hors cuve." Revue Générale Nucléaire, no. 1 (January 1997): 38–43. http://dx.doi.org/10.1051/rgn/19971038.

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37

Sudreau, F., and G. Cognet. "Corium viscosity modelling above liquidus temperature." Nuclear Engineering and Design 178, no. 3 (December 1997): 269–77. http://dx.doi.org/10.1016/s0029-5493(97)00137-4.

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38

Guidez, Joel, Antoine Gerschenfeld, Janos Bodi, Konstantin Mikityuk, Francisco Alvarez-Velarde, Pablo Romojaro, and U. Diaz-Chiron. "ESFR SMART PROJECT CONCEPTUAL DESIGN OF IN-VESSEL CORE CATCHER." EPJ Web of Conferences 247 (2021): 01002. http://dx.doi.org/10.1051/epjconf/202124701002.

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Even before Fukushima accident occurred, the safety authorities have required that new power plant designs must take into account beyond design-basis accidents including possible core meltdown. Among the mitigation strategies, the corium retention must be ensured, so a core catcher is implemented in the design of the Generation IV Sodium-cooled Fast Reactor. An internal core catcher within the vessel (in-vessel retention) is the option chosen for the European Sodium-cooled Fast Reactor investigated in the H2020 ESFR-SMART project. The new core investigated in ESFR SMART with lower void effect has a better behavior in case of severe accident. The use of passive control rods is also an improvement for prevention of severe accident. Moreover, we have in the ESFR SMART core dedicated tubes for corium discharge that should allow discharging quickly the melted materials and should help to prevent large criticality. Calculations show that after several seconds, these discharge tubes begin to open, and the corium arrives by this preferential way on the core catcher, quicker and in limited quantities at the beginning of the accident. However, the core catcher is designed to be able to retain the whole core meltdown. Its design allows good possibilities of cooling by natural convection of sodium. Some thermal calculations were provided with a multi-layer concept but the global mechanical conception seems difficult. So a one layer core catcher in molybdenum, material compatible with sodium and used on the core catcher of the last SFR, started in 2016: BN 800, is investigated. Explanations are given on the choice of this material proposed for the catcher and used for thermal calculations. With the proposed design, the corium is spread on the core catcher and the residual power of the corium can be dispelled by natural convection by the sodium circulating around and above the core catcher without boiling of sodium if the melted core is less than about 25% of whole core. In case of bigger quantities of melted core, boiling of sodium could appear under the core catcher. Further less conservative calculations would be necessary to better know the limit.

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Berge, L., N. Estre, D. Tisseur, E. Payan, D. Eck, V. Bouyer, N. Cassiaut-Louis, C. Journeau, R. Le Tellier, and E. Pluyette. "Fast high-energy X-ray imaging for Severe Accidents experiments on the future PLINIUS-2 platform." EPJ Web of Conferences 170 (2018): 08003. http://dx.doi.org/10.1051/epjconf/201817008003.

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The future PLINIUS-2 platform of CEA Cadarache will be dedicated to the study of corium interactions in severe nuclear accidents, and will host innovative large-scale experiments. The Nuclear Measurement Laboratory of CEA Cadarache is in charge of real-time high-energy X-ray imaging set-ups, for the study of the corium-water and corium-sodium interaction, and of the corium stratification process. Imaging such large and high-density objects requires a 15 MeV linear electron accelerator coupled to a tungsten target creating a high-energy Bremsstrahlung X-ray flux, with corresponding dose rate about 100 Gy/min at 1 m. The signal is detected by phosphor screens coupled to high-framerate scientific CMOS cameras. The imaging set-up is established using an experimentally-validated home-made simulation software (MODHERATO). The code computes quantitative radiographic signals from the description of the source, object geometry and composition, detector, and geometrical configuration (magnification factor, etc.). It accounts for several noise sources (photonic and electronic noises, swank and readout noise), and for image blur due to the source spot-size and to the detector unsharpness. In a view to PLINIUS-2, the simulation has been improved to account for the scattered flux, which is expected to be significant. The paper presents the scattered flux calculation using the MCNP transport code, and its integration into the MODHERATO simulation. Then the validation of the improved simulation is presented, through confrontation to real measurement images taken on a small-scale equivalent set-up on the PLINIUS platform. Excellent agreement is achieved. This improved simulation is therefore being used to design the PLINIUS-2 imaging set-ups (source, detectors, cameras, etc.).

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Kim, Sang Ho, Seong-Wan Hong, and Rae-Joon Park. "Analysis of Steam Explosion under Conditions of Partially Flooded Cavity and Submerged Reactor Vessel." Science and Technology of Nuclear Installations 2018 (July 5, 2018): 1–12. http://dx.doi.org/10.1155/2018/3106039.

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A steam explosion in a reactor cavity makes a mechanical load of the pressure pulse, which can result in a failure of the containment isolation. To prove the integrity of the containment during the ex-vessel steam explosion, the effects of water conditions on a steam explosion have to be identified, and the impulse of a steam explosion has to be exactly assessed. In this study, the analyses for steam explosions were performed for the conditions of a partially flooded cavity and a submerged-vessel in a pressurized water reactor. The entry velocity of a corium jet for the scale of the test facility was varied to simulate the two plant conditions. The TEXAS-V code was used for simulating the phases of premixing and explosion, and the load of a steam explosion was estimated based on the pressure variation. The impulse of a steam explosion under the condition of a corium jet falling into water without a free-fall height is bigger than that under a free-fall height. The fragmented mass of corium in an explosion phase and the distribution of steam fraction are the main parameters for the total load of the steam explosion. This study is expected to contribute to analyses of a steam explosion for a severe accident management strategy.

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Viot, L., R. Le Tellier, and M. Peybernes. "Modeling of the corium crust of a stratified corium pool during severe accidents in light water reactors." Nuclear Engineering and Design 368 (November 2020): 110816. http://dx.doi.org/10.1016/j.nucengdes.2020.110816.

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42

Hemanth Rao, E., Prabhat Kumar Shukla, G. Venkat Reddy, S. S. Murthy, M. Kumaresan, Sanjay Kumar Das, G. Lydia, D. Ponraju, S. Athmalingam, and B. Venkatraman. "Generation of simulated corium using thermite process." Annals of Nuclear Energy 163 (December 2021): 108558. http://dx.doi.org/10.1016/j.anucene.2021.108558.

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43

Hagrman, Donald L., and Joy L. Rempe. "Corium Oxidation at Temperatures above 2000 K." Nuclear Technology 133, no. 2 (February 2001): 194–212. http://dx.doi.org/10.13182/nt01-a3169.

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Kim, Hwan Yeol, Sang Mo An, Jaehoon Jung, Kwang Soon Ha, and Jin Ho Song. "Corium melt researches at VESTA test facility." Nuclear Engineering and Technology 49, no. 7 (October 2017): 1547–54. http://dx.doi.org/10.1016/j.net.2017.06.013.

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Tarabelli, D., G. Ratel, R. Pélisson, G. Guillard, M. Barnak, and P. Matejovic. "ASTEC application to in-vessel corium retention." Nuclear Engineering and Design 239, no. 7 (July 2009): 1345–53. http://dx.doi.org/10.1016/j.nucengdes.2009.02.021.

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Huhtiniemi, I., H. Hohmann, and D. Magallon. "FCI experiments in the corium/water system." Nuclear Engineering and Design 177, no. 1-3 (December 1997): 339–49. http://dx.doi.org/10.1016/s0029-5493(97)00202-1.

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47

Journeau, Christophe, Jean-Francois Haquet, Bertrand Spindler, Claus Spengler, and Jerzy Foit. "The VULCANO VE-U7 Corium spreading benchmark." Progress in Nuclear Energy 48, no. 3 (April 2006): 215–34. http://dx.doi.org/10.1016/j.pnucene.2005.09.009.

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Delacroix, Jules, Romain Le Tellier, and Pascal Piluso. "Oxygen diffusion in liquid (over)stoichiometric corium." Nuclear Engineering and Design 337 (October 2018): 148–60. http://dx.doi.org/10.1016/j.nucengdes.2018.06.027.

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Lomperski, S., and M. T. Farmer. "Performance testing of engineered corium cooling systems." Nuclear Engineering and Design 243 (February 2012): 311–20. http://dx.doi.org/10.1016/j.nucengdes.2011.11.010.

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Levy, Salomon. "Heat transfer during molten corium-concrete interactions." Nuclear Engineering and Design 151, no. 1 (November 1994): 235–46. http://dx.doi.org/10.1016/0029-5493(94)90045-0.

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