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Journal articles on the topic 'Corium'

Author: Grafiati

Published: 4 June 2021

Last updated: 26 July 2025

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1

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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2

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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3

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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4

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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5

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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6

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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7

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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8

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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9

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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<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

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10

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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11

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) (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

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12

Smirnov, Anton D., Ekaterina V. Bogdanova, Pavel A. Pugachev, et al. "Neutronic modeling of a subcritical system with corium particles and water (from international benchmark)." Nuclear Energy and Technology 6, no. 3 (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 irr

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Smirnov, Anton D., Ekaterina V. Bogdanova, Pavel A. Pugachev, et al. "Neutronic modeling of a subcritical system with corium particles and water (from international benchmark)." Nuclear Energy and Technology 6, no. (3) (2020): 155–60. https://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 irr

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14

Gong, Yaopeng, Li Zhang, Yidan Yuan, and Weimin Ma. "Non-Contact Thermophysical Property Measurements of High-Temperature Corium Through Aerodynamic Levitation." Energies 18, no. 1 (2025): 136. https://doi.org/10.3390/en18010136.

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The thermophysical properties of corium are critical for improving the predictive accuracy of severe accident analysis codes. However, due to the high melting temperature and high volatility of corium, thermophysical property measurements are extremely challenging, resulting in a significant lack of data. This study presents a non-contact measurement facility based on the aerodynamic levitation technique, enabling the measurement of the density, surface tension, and viscosity of corium components at temperatures exceeding 3000 K. Density is measured based on the axisymmetric ellipsoid assumpti

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15

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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16

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

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17

Toleubekov, K. O., A. S. Akaev, and M. K. Bekmuldin. "IMPROVING THE EFFICIENCY OF THE INDUCTION HEATING SYSTEM FOR IMITATION THE RESIDUAL ENERGY RELEASE IN THE CORIUM DURING THE INTERACTION WITH HEAT-RESISTANT MATERIALS." NNC RK Bulletin, no. 4 (December 30, 2020): 47–52. http://dx.doi.org/10.52676/1729-7885-2020-4-47-52.

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One of the features of corium is the presence of residual energy release. Induction heating is used as a method for physical modeling of residual energy release at the LAVA-B installation. The article is devoted to the induction heating of corium during experimental studies of its interaction with heat-resistant materials at the Lava-B installation. The results of the analysis of parameters that affect the efficiency of the induction heating system, and determine the optimal conditions for increasing the power and efficiency of simulation of residual energy release are presented. As a result o

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18

Ryzhov, S. N., E. V. Bogdanova, A. A. Ryzhkov, et al. "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 t

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19

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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20

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 explo

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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 (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 sh

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22

Borovoy, A., S. Gavrilov, and V. Hvoschinskiy. "Fukushima. The Danger of Re-Criticality of Destroyed Fuel." ANRI, no. 4 (December 19, 2024): 43–58. https://doi.org/10.37414/2075-1338-2024-119-4-43-58.

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During the accident at three units of the Fukushima-1 Nuclear Power Plant (NPP) the melt of the core materials (corium) destroyed the lower part of the reactor vessels, and some of the corium spilled onto the structures in the containment. As a result, accumulations of FuelContaining Materials (FCM) were formed. In the event of re-criticality both the corium remaining in the reactor vessels and FCM may pose a hazard. Thus when extracting them from damaged NPP units, such hazards must be dully accounted for. The paper focuses on nuclear safety measures both taken and planned in the elimination

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23

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 (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 co

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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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25

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

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

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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 (2000): 107–15. http://dx.doi.org/10.1016/s0029-5493(00)00238-7.

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

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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 (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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Skakov, M. K., N. Ye Mukhamedov, A. V. Pakhnits, and I. I. Deryavko. "NUCLEAR REACTOR CORIUM PROPERTIES OBTAINED AT IGR RESEARCH REACTOR." NNC RK Bulletin, no. 1 (March 30, 2019): 129–32. http://dx.doi.org/10.52676/1729-7885-2019-1-129-132.

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In the paper for the first time thermophysical properties (specific heat capacity, thermal diffusivity and thermal conductivity) of natural corium of a fast nuclear power reactor were determined in the temperature range from room one up to ~400 °С. The obtained data is oriented at use in temperature field calculations when modeling the processes of corium melt retention in fast nuclear reactor vessel.

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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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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 (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 signif

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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 (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

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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, et al. "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 foc

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

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38

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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39

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

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Skakov, Mazhyn, Viktor Baklanov, Gulnur Nurpaissova, Kuanyshbek Toleubekov, Assan Akayev, and Maxat Bekmuldin. "EFFECIENCY OF SIMULATING DECAY HEATIN THE CORIUM OF A NUCLEAR REACTOR BY THE OHMIC HEATING METHOD." Recent Contributions to Physics 90, no. 3 (2024): 108–15. http://dx.doi.org/10.26577/rcph.2024v90i3-013.

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Most commonly, in consequence of a severe accident at a nuclear power plant(NPP) followedby a meltdown of the reactor core, corium is formed. Decay heat is one of the features of it andhighly affectsthe thermal field of the entire corium. Thus, takinginto account the decay heat is a central tophysical modelingof severe accidents. For this reason, methods for simulating decay heatmust meet at least two requirements of physical modeling: intensity and uniformity of heating throughout the entire volume of the melt.This paper provides studies on the efficiencyof using ohmic heating as a method for

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García-Lascuráin, Alma A., Gabriela Aranda-Contreras, Margarita Gomez-Chavarin, et al. "Tratamiento de la laminitis crónica en equinos utilizando células troncales mesenquimales alogénicas de la médula ósea." Revista Mexicana de Ciencias Pecuarias 12, no. 3 (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 physiopath

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Tisseur, D., M. Cavaro, F. Rey, et al. "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. Furth

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Kalvand, Ali, and I. V. Kazachkov. "Peculiarities of the melting-solidification processes by sinking of the melting blocks into high-temperature corium melt." Nuclear Physics and Atomic Energy 10, no. 2 (2009): 178–84. https://doi.org/10.15407/jnpae2009.02.178.

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Mathematical modeling of the corium melt coolability with the low-temperature melting blocks is a subject of the paper, focusing on the methodology of modeling. The results of a computer simulation and conclusions concerning the features of the process studied are presented. Obtained results allowed making some conclusions concerning the main features and dynamics of the process of corium melt pool cooling with sinking of low-melting temperature blocks into high-temperature pool, which present the key part of the passive protection system against severe accidents at NPP.

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Konoval, O. V., Ali Kalvand, and I. V. Kazachkov. "Modeling of the corium cooling and loading factor analysis for containment during severe accidents." Nuclear Physics and Atomic Energy 14, no. 3 (2013): 276–87. https://doi.org/10.15407/jnpae2013.03.276.

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The paper is devoted to the development and study of the mathematical model for corium melt interaction with low-temperature melting blocks in the passive protection systems (PPS) against severe accidents at the NPP, and learning the peculiarities of construction and operation of the PPS. The configurations of cooling blocks' distributions considered and the results of their work in the corium cooling pool are compared to the data of other PPS's conceptions. The conclusion is made that the models developed and the results obtained may be useful for constructing the PPS against severe accidents

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Skakov, M. К., K. O. Toleubekov, M. K. Bekmuldin, and A. S. Akaev. "DEVELOPMENT OF A THERMOPHYSICAL MODEL FOR THE EXPERIMENTAL ASSEMBLY OF THE VCG-135 TEST BENCH TO STUDY THE INTERACTION OF CORIUM WITH METAL-COOLER IN THE CONDITIONS OF A SEVERE ACCIDENT." NNC RK Bulletin, no. 4 (December 30, 2024): 5–11. https://doi.org/10.52676/1729-7885-2024-4-5-11.

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This article presents the results of modeling of the temperature field of the experimental assembly of the VCG-135 test bench to study the interaction between model corium and candidate metal-coolers (zinc, antimony and manganese) in the conditions of a severe accident.The need for modeling is associated with the probability of metal melting in the discharge device due to the heat flow from the heating crucible of the experimental assembly. Thus, the purpose of the modeling was the justification of the integrity of the design of the metal discharge device during the production of liquid corium

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Moghaddam, V. H., and I. V. Kazachkov. "Modeling of the corium jet penetration into the pool of volatile coolant under reactor vessel." Nuclear Physics and Atomic Energy 11, no. 2 (2010): 151–58. https://doi.org/10.15407/jnpae2010.02.151.

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Developed mathematical model and some results of its application to the element of the passive protection system against severe accidents at NPP with water pool under reactor vessel for corium melt and particles' cooling are presented. This element of a system is penetration behavior of the high-temperature corium jet into volatile coolant pool, with account of the vapor produced on a melt jet penetration. Success of the problem solution determines an effectiveness and durability of the passive protection system against severe accidents. Therefore, the obtained results may be useful for the de

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Guidez, Joel, Antoine Gerschenfeld, Janos Bodi, et al. "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

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Berge, L., N. Estre, D. Tisseur, et al. "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 ab

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Sunil, Kumar Jatav, Kumar Pandey Vijay, Pandel U., K. Nayak A., and Kumar Duchaniya Rajendra. "Thermo-Physical Properties of CaO-Fe2O3 Binary Mixture and its Application in the Field of Nuclear Reactor as Simulant Material." International Journal of Engineering and Advanced Technology (IJEAT) 9, no. 3 (2020): 1706–9. https://doi.org/10.35940/ijeat.C5549.029320.

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The simulant materials play important role in the melt coolability experiments to understand the actual scenarios of core melt accidents in the field of nuclear reactor. Simulant materials are generally oxide/ceramics materials which have the properties similar to the properties of corium (mixture of UO2 , ZrO2 , Zr alloy, Fe, Ni and Cr etc.). This work was carried out to determine the thermo-physical properties of CaO-Fe2O3 binary mixtures of different ratio of CaO and Fe2O3 (23C77F, 26C74F, 29C71F, 32CF68 and 35C65F; here the ratio is in the wt% and C for CaO and F for Fe2O3 ) and compare th

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