FCI experiments in the corium/water system

The KROTOS fuel coolant interaction (FCI) tests are aimed at providing benchmark data to examine the effect of fuel/coolant initial conditions and mixing on explosion energetics. Experiments, fundamental in nature, are performed in well-controlled geometries and are complementary to the FARO large scale tests. Recently, a test series was performed using 3 kg of prototypical corium (80 w/o UO₂, 20 w/o ZrO₂) which was poured into a water column of ≤1.25 m in height (95 and 200 mm in diameter) under 0.1 MPa ambient pressure. Four tests were performed in the test section of 95 mm in diameter (ID) with different subcooling levels (10–80 K) and with and without an external trigger. Additionally, one test has been performed with a test section of 200 mm in diameter (ID) and with an external trigger. No spontaneous or triggered energetic FCIs (steam explosions) were observed in these corium tests. This is in sharp contrast with the steam explosions observed in the previously reported alumina (Al₂O₃) test series which had the same initial conditions of ambient pressure and subcooling. The post-test analysis of the corium experiments indicated that strong vaporisation at the melt/water contact led to a partial expulsion of the melt from the test section into the pressure vessel. In order to avoid this and to obtain a good penetration and premixing of the corium melt, an additional test was performed with a larger diameter test section. In all the corium tests an efficient quenching process (0.8–1.0 MW kg-melt⁻¹) with total fuel fragmentation (mass mean diameter 1.4–2.5 mm) was observed. Results from alumina tests under the same initial conditions are also given to highlight the differences in behaviour between corium and alumina melts during the melt/water mixing.     

Water boiling on the corium melt surface under VVER severe accident conditions

Experimental results are presented on the interaction of corium melt with water supplied on its surface. The tests were conducted in the ‘Rasplav-2’ experimental facility. Corium melt was generated by induction melting in the cold crucible. The following data were obtained: heat transfer at boiling water-melt surface interaction, gas and aerosol release, post-interaction solidified corium structure. The corium melt charge had the following composition, mass%: 60% UO_2+x–16% ZrO₂–15% Fe₂O₃–6% Cr₂O₃–3% Ni₂O₃. The melt surface temperature ranged within 1920–1970 K.     

Results of recent KROTOS FCI tests: alumina versus corium melts

Recent results from KROTOS fuel-coolant interaction experiments are discussed. Five tests with alumina were performed under highly subcooled conditions, all of these tests resulted in spontaneous steam explosions. Additionally, four tests were performed at low subcooling to confirm, on one hand, the suppression of spontaneous steam explosions under such conditions and, on the other hand, that such a system is still triggerable using an external initiator. The other test parameters in these alumina tests included the melt superheat and the initial pressure. All the tests in the investigated superheat range (150–750 K) produced a steam explosion and no evidence of the explosion suppression by the elevated initial pressure (in the limited range of 0.1–0.375 MPa) was observed in the alumina tests. The corium test series include a test with 3 kg of melt under both subcooled and near saturated conditions at ambient pressure. Two additional tests were performed with subcooled water; one test was performed at an elevated pressure of 0.2 MPa with 2.4 kg of melt and another test with 5.1 kg of melt at ambient pressure. None of these tests with corium produced a propagating energetic steam explosion. However, propagating low energy (about twice the energy of the trigger pulse) events were observed. All corium tests produced significantly higher water level swells during the mixing phase than the corresponding alumina tests. Present experimental evidence suggests that the water depletion in the mixing zone suppresses energetic steam explosions with corium melts at ambient pressure and in the present pour geometry. Processes that could produce such a difference in void generation are discussed.     

The effect of corium composition and interaction vessel geometry on the prototypic steam explosion

This paper discusses the results of steam explosion experiments using molten material consisting of UO₂ and ZrO₂ mixture, which is called corium, to simulate a prototypic steam explosion in a nuclear reactor during a postulated severe accident. About 5–10 kg of molten material with enough superheat was poured into a pool of water in a test section at room temperature to simulate ex-vessel steam explosion in the reactor situation. Most of the experiments were externally triggered. The purpose of the experiments was to investigate the effect of material composition and average void fraction on the strength of a prototypic steam explosion, which were highlighted as major unresolved issues.The experiments were performed using two kinds of mixtures, one, corium A, at 70:30 weight percent composition of UO₂ and ZrO₂, close to eutectic composition, and the other, corium B, at 80:20 weight percent. Also, two kinds of cylindrical test sections having a different diameter were used. It turned out that corium A was likely to produce an energetic steam explosion, while corium B seldom led to an energetic steam explosion. The existence of mush phase for the non-eutectic mixture is suggested to be the reason for the difference. Comparative cross sectional views of the corium particles by scanning electron microscope supported the proposed argument. The tests performed with a narrow test section seldom led to an energetic steam explosion for both materials. An increase in average void is suggested to be the reason for the non-explosive behavior, which is consistent with the physical models employed in the current steam explosion computer codes.     

A use of prototypic material for the investigation of severe accident progression

An experimental research platform using corium melts is established for the understanding of safety related important phenomena during a severe accident progression. The research platform includes TROI facility for corium water interaction experiments and VESTA facility for corium-structural material interaction experiments. A cold crucible technology is adapted and improved for a generation of 5–100 kg of corium melts at various compositions. TROI facility is used for experiments to investigate premixing and explosion behaviors during a fuel coolant interaction process. More than 70 experiments using corium at various compositions were performed to simulate steam explosion phenomena in a reactor situation. The results indicate that the conversion efficiency of steam explosion for corium is less than 1%. VESTA facility is used to investigate molten corium-structural material interaction phenomena. VESTA facility consists of two cold crucibles. One crucible is used for the melting of charged material and pouring of corium melt. The other crucible is used for the corium-structural material interaction while providing an induction heating to simulate the decay heat. The results of an experiment on the interaction between corium melt and a specimen made of Inconel performed in the VESTA facility is reported.     

Phase separation of metal-added corium and its effect on a steam explosion

To simulate a relocation of molten core material and its interaction phenomenon with water during a severe accident in a nuclear reactor, a typical corium of UO₂/ZrO₂/Zr/Stainless steel mixed at a 62 wt%, 15 wt%, 12 wt% and 11 wt%, respectively, was melted and then cooled down to become a solidified ingot. It was shown that the molten corium was separated into two layers, of which the upper layer was oxide mixtures and the lower layer was metal alloys. The upper layer was UO₂ and ZrO₂ and the lower layer mostly consisted of metal mixtures such as uranium, zirconium and stainless steel. Iron content varied with the positions and about a half of it existed as an alloy such as Fe₂U. Uranium metal was produced by reduction of UO₂ by zirconium metal. The average densities of the upper oxide layer and the lower metal layer were 8.802 and 9.411 g/cm³, respectively. In another test, metal-added molten corium was poured into water and it showed that a steam explosion could occur by applying an external trigger.     

Fuel coolant interaction experiments in TROI using a UO2/ZrO2 mixture

This paper discusses the results of steam explosion experiments using reactor material carried out under “Test for Real cOrium Interaction with water (TROI)” program. About 4–9 kg of corium melt jet is delivered into a sub-cooled water pool at atmospheric pressure. Spontaneous steam explosions are observed in four tests among six tests. The dynamic pressure, dynamic load, and morphology of debris clearly indicate the cases with steam explosion. The initial conditions and results of the experiments are discussed.     

Effect of melt water interaction configuration on the process of steam explosion

Steam explosion experiments are performed at various modes of melt water interaction configuration using prototypic corium melt. The tests are performed to simulate both melt water interaction in a partially flooded cavity and melt water interaction in a cavity with submerged reactor. The tests are performed using zirconia and corium melts. The behavior of melt jet fragmentation during the flight in the air and fragmentation and mixing of melt jet in water is investigated by a high-speed video visualization and by comparison of debris size distribution and morphology of debris. Strength of steam explosion is estimated by measuring dynamic pressure and dynamic force.     

Experimental investigation of corium quenching at elevated pressure

This paper describes the results of experiments designed to quantify the cooling rate of corium by an overlying water pool. The experiments are intended to provide fundamental information on the ability of water to ingress into cracks and fissures that form in the debris during quench, thereby augmenting the otherwise conduction-limited heat transfer process. This information is being used to assess the effectiveness of a water pool in thermally stabilizing a molten-core/concrete interaction and cooling of ex-vessel core debris. The experiments involved corium inventories of 75 kg with a melt depth of 15 cm and diameter of 30 cm. The corium was composed of UO₂/ZrO₂/concrete to simulate mixtures of molten reactor core components and either siliceous or limestone/common sand (LCS) concrete. Initial melt temperatures were of the order of 2100 °C. The heat transfer rate from the corium was determined through measurements of the vapor production rate from the water pool. The melt was quenched at atmospheric pressure for the first two tests and at 4 bar for the two subsequent tests. Preliminary data analysis indicates that the overall heat transfer rate exceeded the conduction-limited rate for the three melts containing 8 wt.% concrete, but not for the fourth, which had 23 wt.% concrete. Also, the quench rate of the 8 wt.% concrete melts did not vary appreciably with pressure.     

Characteristics of corium debris bed generated in large-scale fuel-coolant interaction experiments

The formation of corium debris as the result of fuel-coolant interaction (energetic or not) has been studied experimentally in the FARO and KROTOS facilities operated at JRC-Ispra between 1991 and 1999. Experiments were performed with 3–177 kg of UO₂–ZrO₂ and UO₂–ZrO₂–Zr melts, quenched in water at depth between 1 and 2 m, and pressure between 0.1 and 5.0 MPa. The effect of various parameters such as melt composition, system pressure, water depth and subcooling on the quenching processes, debris characteristics and thermal load on bottom head were investigated, thus, giving a large palette of data for realistic reactor situations.Available data related to debris coolability aspects in particular are:

• Geometrical configuration of the collected debris.

• Partition between loose and agglomerated (“cake”) debris.

• Particle size distribution with and without energetic interaction.

These data are synthesised in the present contribution.     

Corium crust strength measurements

Corium strength is of interest in the context of a severe reactor accident in which molten core material melts through the reactor vessel and collects on the containment basemat. Some accident management strategies involve pouring water over the melt to solidify it and halt corium/concrete interactions. The effectiveness of this method could be influenced by the strength of the corium crust at the interface between the melt and coolant. A strong, coherent crust anchored to the containment walls could allow the yet-molten corium to fall away from the crust as it erodes the basemat, thereby thermally decoupling the melt from the coolant and sharply reducing the cooling rate. This paper presents a diverse collection of measurements of the mechanical strength of corium. The data is based on load tests of corium samples in three different contexts: (1) small blocks cut from the debris of the large-scale MACE experiments, (2) 30 cm-diameter, 75 kg ingots produced by SSWICS quench tests, and (3) high temperature crusts loaded during large-scale corium/concrete interaction (CCI) tests. In every case the corium consisted of varying proportions of UO₂, ZrO₂, and the constituents of concrete to represent a LWR melt at different stages of a molten core/concrete interaction. The collection of data was used to assess the strength and stability of an anchored, plant-scale crust. The results indicate that such a crust is likely to be too weak to support itself above the melt. It is therefore improbable that an anchored crust configuration could persist and the melt become thermally decoupled from the water layer to restrict cooling and prolong an attack of the reactor cavity concrete.     

Corium viscosity modelling above liquidus temperature

In the event of a highly unlikely core melt-down accident in Pressurized Water Reactors (PWR), scenarios in which the reactor pressure vessel fails and the core melt mixture (called corium) relocates into the reactor cavity, cannot be excluded. The Nuclear Reactor Division (CEA/DRN) has undertaken investigations in order to model rheological corium behaviour. In this paper, a bibliographic study and a comparison with available data lead to the conclusion that the viscosity of corium containing UO₂, ZrO₂ and Zr can be calculated, at the melting point, with a good accuracy using the Andrade formula. Above the liquidus temperature, the correlations, proposed for pure metals and metal alloys, between the activation energy and the melting temperature are not available in the case of urania and thus cannot be used to calculate liquid corium activation energy. To overcome this problem, in this paper we propose to use a mole fraction averaged activation energy. Nevertheless, this last point needs to be validated.     

Ex-vessel fuel-coolant interaction experiment in the DISCO facility in the LACOMECO project

In the frame of the LACOMECO (large scale experiments on core degradation, melt retention and containment behavior) project of the 7th European Framework Program, a test in the DISCO (dispersion of corium) facility was performed in order to analyze the phenomena which occur during an ex-vessel fuel–coolant interaction (FCI). The test is focused on the premixing phase of the FCI with no trigger used for explosion phase. The objectives of the test were to provide data concerning the dispersion of water and melt out of the pit, characterization of the debris and pressurization of the reactor compartments for scenarios, where the melt is ejected from the reactor pressure vessel (RPV) under pressure. The experiment was performed for a reactor pit geometry close to a French 900 MWe reactor configuration at a scale of 1:10. The corium melt was simulated by a melt of iron–alumina with a temperature of 2400 K. A containment pressure increase of 0.04 MPa was measured, the total pressure reached about 0.24 MPa. No spontaneous steam explosion was observed. About 16% of the initial melt (11.62 kg) remained in the RPV vessel, 60% remained in the cavity mainly as a compact crust. The fraction of the melt transported out of the pit was about 24%.     

Experimental evaluation of the water ingression mechanism for corium cooling

Experiments were performed to assess the significance of water ingression cooling in the quenching of molten corium. Water ingression is a mechanism by which water penetrates into cracks and pores of solidified corium to enhance cooling that would otherwise be severely limited by the low thermal conductivity of the material. Quench tests were conducted with 2100 °C melts weighing 75 kg composed of UO₂, ZrO₂ and chemical constituents of concrete. The amount of concrete in the melts was varied between 4% and 23%. The melts were quenched with an overlying water layer; three tests were conducted at a system pressure of 1 bar and four tests at 4 bar. The measured cooling rates were found to decrease with increasing concrete content and, contrary to expectations, are essentially independent of system pressure. For the lower concrete content melts, cooling rates exceeded the conduction-limited rate with the difference being attributed to the water ingression mechanism. Measurements of the permeability of the corium “ingots” produced by the quench tests were used to obtain a second, independent set of dryout heat flux data, which exhibits the same trend as the quench test data. The data was used to validate an existing dryout heat flux model based on corium permeability associated with thermally induced cracking. The model uses the thermal and mechanical properties of the corium and coolant, and it reproduces the very particular data trend found for the dryout heat flux as a function of concrete content. The model predicts that water ingression cooling would be most effective for concrete-free corium mixtures such as in-vessel type melts. For such a melt the model predicts a dryout heat flux of 400 kW/m² at a pressure of 1 bar. The results of this study provide an experimental basis for a water ingression model that can be incorporated into computer codes used to assess accident management strategies.     

Corium kinematic viscosity measurement

Viscosity is one of the important properties which along with thermal conductivity and coefficient of volumetric expansion, determines convection in molten corium in the case of unfavourable progress of a severe accident in nuclear reactors. The viscosimetric attachment has been designed, developed and built for the purpose of measurement of corium kinematic viscosity with the technique of torsional oscillations. The created experimental devices and procedure for measurement of liquid corium viscosity have been tested and refined over temperature range up to 2800°C. The experimental data of 62UO₂+38ZrO₂ (mol%) melt viscosity have been obtained at temperature ranging from 2600 to 2800°C. The viscosity of 62UO₂+8ZrO₂+30Zr (mol%) melt has been measured at temperatures from 2400 to 2700°C. This work has been performed within OECD RASPLAV Project. The Management Board approves publication of this paper.     

Ex-vessel corium spreading: results from the VULCANO spreading tests

In the hypothetical case of a nuclear reactor severe accident, the reactor core could melt and form a mixture, called corium, of highly refractory oxides (UO₂, ZrO₂) and metallic or oxidized steel, that could eventually flow out of the vessel and mix with the basemat decomposition products (generally oxides such as SiO₂, Al₂O₃, CaO, Fe₂O₃, …). For some years, the French Atomic Energy Commission (CEA) has launched an R&D program which aimed at providing the tools for improving the mastering of severe accidents.Within this program, the VULCANO experimental facility is operated to perform experiments with prototypic corium (corium of realistic chemical composition including depleted UO₂). This is coupled with the use of specific high-temperature instrumentation requiring in situ cross calibration. This paper is devoted to the “spreading experiments” performed in the VULCANO facility, in which the effects of flow and solidification are studied.Due to the complex behavior of corium in the solidification range, an interdisciplinary approach has been used combining thermodynamics of multicomponent mixtures, rheological models of silicic semisolid materials, heat transfer at high temperatures, free-surface flow of a fluid with temperature-dependant properties.Twelve high-temperature spreading tests have been performed and analyzed. The main experimental results are the good spreadability of corium–concrete mixtures having large solidification ranges even with viscous silicic melts, the change of microstructure due to cooling rates, the occurrence of a large thermal contact resistance at the corium–substrate interface, the presence of a steep viscosity gradient at the surface, the transient concrete ablation. Furthermore, the experiments showed the presence of the gaseous inclusions in the melt even without concrete substrate. This gas release is linked to the local oxygen content in the melt which is function of the nature of the atmosphere, of the phases (FeO_x, UO_y, …) and of the substrate. These tests with prototypic material have improved our knowledge on corium and contributed to validate spreading models and codes which are used for the assessment of corium mastering concepts.     

On the mechanism of aluminum ignition in steam explosions

An available theory [Epstein, M., Fauske, H.K., 1994. A crystallization theory of underwater aluminum ignition. Nucl. Eng. Des. 146, 147–164] of the ignition of aluminum melt drops under water, which is based on the assumption that the aluminum oxide (Al₂O₃) drop-surface skin first appears in a metastable molten state, is compared with existing experimental data on the ignition of aluminum drops behind shock waves in water [Theofanous, T.G., Chen, X., DiPiazza, P., Epstein, M., Fauske, H.K., 1994. Ignition of aluminum droplets behind shock waves in water, Phys. Fluids 6, 3513–3515]. The predicted and measured ignition temperature of about 1770 K coincides approximately with the spontaneous nucleation temperature of supercooled liquid Al₂O₃ (1760 K). This suggests that the crystallization of the oxide layer represents a strong ‘barrier’ to aluminum drop ignition under water. Apparently a similar interpretation is applicable to aluminum drop ignition in gaseous oxidizing atmospheres. We conclude from the theory that the low-temperature aluminum ignitions (in the range 1100–1600 K) that have been observed during steam explosions are a consequence of the short aluminum drop oxidation times in this environment relative to the characteristic time for Al₂O₃ crystallization. Several aspects of the aluminum drop/shock interaction experiments besides ignition are discussed in the paper. In particular, the experiments provide strong evidence that during the course of a vapor explosion metal fragmentation occurs via a thermal mechanism at low pressure and precedes the development of a high-pressure shock.     

High pressure corium melt quenching tests in FARO

First-of-a-kind experimental data on the quenching of large masses of corium melt of realistic composition when poured into pressurised water at reactor scale depths are presented and discussed. The tests involved 18 and 44 kg of a molten mixture 80 w% UO₂-20 w% ZrO₂, which were delivered by gravity through a nozzle of diameter 0.1 m to 1 m depth nearly saturated water at 5.0 MPa. The objective was to gain early information on the melt/water quench process previous to tests that will involve larger masses of melt (1.50 kg of mixtures UO₂---ZrO₂---Zr). Particularly, pressures and temperatures were measured both in the gas phase and in the water. The results show that significant quenching occurred during the melt fall stage with 30% to 42% of the melt energy transferred to the water. About two-thirds of the melt broke up into particles of mean size of the order of 4.0 mm. The remaining one-third collected still molten in the debris catcher but did not produce any damage to the bottom plate. The maximum downward heat flux was 0.8 MW m². The maximum vessel overpressurisation, i.e. 1.8 MPa, was recorded with 44 kg of melt poured into 255 kg of water and a gas phase volume of 0.875 m³. No steam explosions occurred.     

熔融物液柱碎化模型对先进压水堆堆外蒸汽爆炸计算的影响分析

在堆外蒸汽爆炸计算中,液柱碎化模型影响着熔融物液滴生成速率、液滴直径、液滴分布、液滴凝固和气泡比例等粗混合参数和现象,从而影响了蒸汽爆炸的冲击载荷。本文基于MC3D V3.8程序,采用不同的液柱碎化模型（CONST模型和KHF模型）对先进压水堆堆外蒸汽爆炸进行计算分析,探讨了CONST和KHF模型对蒸汽爆炸计算的影响。结果表明,两种模型计算的粗混合状态类似;在熔融物触底时刻,爆炸性准则几乎相同,此时触发爆炸得到的冲击载荷差别很小,表明该时刻触发爆炸时不同液柱碎化模型对爆炸冲击计算的影响较小;在本文所定义的工况下,先进压水堆堆坑墙体承受的最高压力约为20 MPa,最大冲量小于0.2 MPa•s。     

FCI experiments in the aluminum oxide/water system

The KROTOS facility at JRC Ispra was recently used to study experimentally melt-coolant premixing and steam explosion phenomena in Al₂O₃/water mixtures with approximately 1.5 kg melt at 2300–2400 °C. In the five tests performed the main parameter was the water subcooling, 10, 40 and 80 X, respectively. In the nearly saturated system, steam explosions could be externally triggered, which resulted in high (supercritical) explosion pressures in the test tube: KROTOS 26, 28. Without triggering, melt penetration in water and melt agglomeration on the bottom plate of the test tube could be observed, which gave rise to strong steaming during the melt cooling-down process: KROTOS 27. In the two tests KROTOS 29, 30, performed with 80 K subcooled water, self-triggered steam explosions occurred with pressures of more than 100 MPa. Post-test analysis of the debris revealed that 85% of the interacting fuel mass fragmented in particles of sizes smaller than 250 μm. An energy conversion ratio of 1.25% was estimated from vessel pressurization data taking into account the energy content in the fuel mass which fragmented to particle diameters of less than 250 μm. The test section was damaged in the test KROTOS 30.