Particle Size Characteristics of Molten Corium Quenched in Water

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

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Numerical Simulation of Corium Jet Breakup During Fuel-Coolant Interaction Based on the ALISA Test Performed at KROTOS Test Facility

Volume 8: Computational Fluid Dynamics (CFD); Nuclear Education and Public Acceptance ◽

10.1115/icone26-81818 ◽

2018 ◽

Author(s):

Pei Shen ◽

Wenzhong Zhou

Keyword(s):

Numerical Simulation ◽

Steam Explosion ◽

Field Method ◽

Severe Accident ◽

Phase Field Method ◽

Test Facility ◽

Jet Breakup ◽

Jet Behavior ◽

Jet Fragmentation ◽

Explosion Test

Steam explosion is one of the consequences of fuel-coolant interactions in a severe accident. Melt jet fragmentation, which is the key phenomenon during steam explosion, has not been clarified sufficiently which prevents the precise prediction of steam explosion. The focus of this paper is on the numerical simulation of the melt jet behavior falling into a coolant pool in order to get a qualitative and quantitative understanding of initial premixing stage of fuel-coolant interaction. The objective of our first phase is the simulation of the fragmentation process and the estimation of the jet breakup length. A commercial CFD code COMSOL is used for the 2D numerical analysis employing the phase field method. The simulation condition is similar to our steam explosion test supported by the ALISA (Access to Large Infrastructure for Severe Accidents) project between European Union and China, and carried out in the KROTOS test facility at CEA, France. The simulation result is in relatively good agreement with the experimental data. Then the effect of the initial jet velocity, the jet diameter and the instability theory are presented. The preliminary data of melt jet fragmentation is helpful to understand the premixing stage of the fuel-coolant interaction.

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Dominant Factor for the Occurrence of a Steam Explosion

Defect and Diffusion Forum ◽

10.4028/www.scientific.net/ddf.273-276.388 ◽

2008 ◽

Vol 273-276 ◽

pp. 388-393

Author(s):

B.T. Min ◽

S.W. Hong ◽

J.H. Kim ◽

I.K. Park ◽

H.D. Kim

Keyword(s):

Steam Explosion ◽

Nuclear Reactor ◽

Cold Water ◽

Hydrogen Reduction ◽

Chemical Compounds ◽

Hydrogen Gas ◽

Dominant Factor ◽

Small Particles ◽

Debris Particles ◽

Pure Zro2

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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In-Vessel Corium Behavior and Steam Explosion Risks: A Review of Experimental Approaches

Volume 7: Fuel Cycle, Decontamination and Decommissioning, Radiation Protection, Shielding, and Waste Management; Mitigation Strategies for Beyond Design Basis Events ◽

10.1115/icone25-67176 ◽

2017 ◽

Author(s):

Pei Shen ◽

Wenzhong Zhou

Keyword(s):

Heat Transfer ◽

Steam Explosion ◽

Large Scale ◽

Nuclear Reactor ◽

Reactor Core ◽

Reactor Vessel ◽

Radioactive Material ◽

Severe Accident ◽

Reactor Accident ◽

Experimental Approaches

Although no one would like to see, a severe nuclear reactor accident may result in reactor core melting, the fuel melt dropping into water in the reactor vessel, and then interacting with coolant into steam explosion. Steam explosion is a result of very rapid and intense heat transfer and violent interaction between the high temperature melt and low temperature coolant. The timescale for heat transfer is shorter than that for pressure relief, resulting in the formation of shock waves and/or the production of missiles at a later time during the expansion of coolant steam explosion. Steam explosion may endanger the reactor vessel and surrounding structures. During a severe reactor accident scenario, steam explosion is an important risk, even though its probability to occur is pretty low, since it could lead to large releases of radioactive material, and destroy the containment integrity. This study provides a comprehensive review of vapor explosion experiments, especially the most recent ones. In this review, fist, small to intermediate scale experiments related to premixing, triggering and propagation stages are reviewed and summarized in tables. Then the intermediate to large scale experiments using prototypic melt are reviewed and summarized. The recent OECD/SERENA2 project including KROTOS and TROI facilities’ work is also discussed. The studies on steam explosion are vital for reactor severe accident management, and will lead to improved reactor safety.

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Analysis of Steam Explosion under Conditions of Partially Flooded Cavity and Submerged Reactor Vessel

Science and Technology of Nuclear Installations ◽

10.1155/2018/3106039 ◽

2018 ◽

Vol 2018 ◽

pp. 1-12

Author(s):

Sang Ho Kim ◽

Seong-Wan Hong ◽

Rae-Joon Park

Keyword(s):

Steam Explosion ◽

Mechanical Load ◽

Free Fall ◽

Severe Accident ◽

Test Facility ◽

Pressurized Water Reactor ◽

Pressurized Water ◽

Total Load ◽

Entry Velocity ◽

Fall Height

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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Assessment of the MARS Code Using the Two-Phase Natural Circulation Experiments at a Core Catcher Test Facility

Science and Technology of Nuclear Installations ◽

10.1155/2017/5731420 ◽

2017 ◽

Vol 2017 ◽

pp. 1-13

Author(s):

Dong Hun Lee ◽

Su Ryong Choi ◽

Kwang Soon Ha ◽

Han Young Yoon ◽

Jae Jun Jeong

Keyword(s):

Computational Analysis ◽

Nuclear Reactor ◽

Natural Circulation ◽

Three Dimensional ◽

Heating Surface ◽

Severe Accident ◽

Test Facility ◽

Two Phase ◽

Core Catcher ◽

Parametric Effects

A core catcher has been developed to maintain the integrity of nuclear reactor containment from molten corium during a severe accident. It uses a two-phase natural circulation for cooling molten corium. Flow in a typical core catcher is unique because (i) it has an inclined cooling channel with downwards-facing heating surface, of which flow processes are not fully exploited, (ii) it is usually exposed to a low-pressure condition, where phase change causes dramatic changes in the flow, and (iii) the effects of a multidimensional flow are very large in the upper part of the core catcher. These features make computational analysis more difficult. In this study, the MARS code is assessed using the two-phase natural circulation experiments that had been conducted at the CE-PECS facility to verify the cooling performance of a core catcher. The code is a system-scale thermal-hydraulic (TH) code and has a multidimensional TH component. The facility was modeled by using both one- and three-dimensional components. Six experiments at the facility were selected to investigate the parametric effects of heat flux, pressure, and form loss. The results show that MARS can predict the two-phase flow at the facility reasonably well. However, some limitations are obviously revealed.

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