Experiments on the interactions of molten ZrO2 with water using TROI facility

Korea Atomic Energy Research Institute (KAERI) launched an intermediate scale steam explosion experiment named ‘Test for Real cOrium Interaction with water (TROI)’ using reactor material. The objective of the program is to investigate whether the corium would lead to energetic steam explosion when interacted with cold water at a low pressure. The melt/water interaction is made in a multi-dimensional test section located in a pressure vessel. The inductive skull melting, which is basically a direct inductive heating of an electrically conducting melt, is implemented for the melting and delivery of corium. In the first series of tests using several kg of ZrO₂ where the melt/water interaction is made in a heated water pool at 30–95 °C, either a quenching or a spontaneous steam explosion was observed. The spontaneous explosion observed in the present ZrO₂ melt/water experiments clearly indicates that the physical properties of the UO₂/ZrO₂ mixture have a strong effect on the energetics of steam explosion.     

Energy Transformations During the Thermal Interaction of Sodium with Core Materials Melt

The results of experimental studies of the thermal interaction of core materials melt using simulators (UO₂ + Mo, ZrO₂ + Fe melts at 3000–3100 K) and sodium at 823 K are presented.The kinematic characteristics of the displacement of sodium during the interaction are estimated on the basis of two methods used simultaneously. The first method is based on measuring the residual flexure strain of precalibrated plate-shaped components, which are located above the sodium level, in a plane perpendicular to the axis of the interaction chamber. In the second method, the sodium consumption is measured in a pipe which bypasses the interaction chamber. The data obtained are used to calculate the work of irreversible adiabatic compression of gas (argon) in the cavity of the bypass pipe conduit.The conversion ratios for the conversion of the thermal energy of the melt into the work of expansion of the system are 5·10^–3–10^–2%. The two methods are shown to agree satisfactorily with one another.A model of the thermal interaction of the melt with sodium is proposed on the basis of the experimental results and correlation of existing data.     

Behavior of molybdenum in pyrochemical reprocessing: A spectroscopic study of the chlorination of molybdenum and its oxides in chloride melts

The high temperature reactions of molybdenum and its oxides with chlorine and hydrogen chloride in molten alkali metal chlorides were investigated between 400 and 700 °C. The melts studied were LiCl-KCl, NaCl-CsCl and NaCl-KCl and the reactions were followed by in situ electronic absorption spectroscopy measurements. In these melts Mo reacts with Cl₂ and initially produces MoCl₆²⁻ and then a mixture of Mo(III) and Mo(V) chlorocomplexes, the final proportion depending on the reaction conditions. The Mo(V) content can be removed as MoCl₅ from the melt under vacuum or be reduced to Mo(III) by Mo metal. The reaction of Mo when HCl gas is bubbled into alkali chloride melts yields only MoCl₆³⁻. MoO₂ reacts in these melts with chlorine to form soluble MoOCl₅²⁻ and volatile MoO₂Cl₂. MoO₃ is soluble in chloride melts and then decomposes into the oxychloride MoO₂Cl₂, which sublimes or can be sparged from the melt, and molybdate. Pyrochemical reprocessing can thus be employed for molybdenum since, after various intermediates, the end-products are chloride melts containing chloro and oxychloro anions of molybdenum plus molybdate, and volatile chlorides and oxychlorides that can be readily separated off. The reactions were fastest in the NaCl-KCl melt. The X-ray diffraction pattern of MoO₂Cl₂ is reported for the first time.     

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.     

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.     

Transformation and fragmentation behavior of molten metal drop in sodium pool

In order to clarify the fragmentation mechanism of a metallic alloy (U–Pu–Zr) fuel on liquid phase formed by metallurgical reactions (liquefaction temperature = 650 °C), which is important in evaluating the sequence of core disruptive accidents for metallic fuel fast reactors, a series of experiments was carried out using molten aluminum (melting point = 660 °C) and sodium mainly under the condition that the boiling of sodium does not occur. When the instantaneous contact interface temperature (T_i) between molten aluminum drop and sodium is lower than the boiling point of sodium (T_c,bp), the molten aluminum drop can be fragmented and the mass median diameter (D_m) of aluminum fragments becomes small with increasing T_i. When T_i is roughly equivalent to or higher than T_c,bp, the fragmentation of aluminum drop is promoted by thermal interaction caused by the boiling of sodium on the surface of the drop. Furthermore, even under the condition that the boiling of sodium does not occur and the solid crust is formed on the surface of the drop, it is confirmed from an analytical evaluation that the thermal fragmentation of molten aluminum drop with solid crust has a potential to be caused by the transient pressurization within the melt confined by the crust. These results indicate the possibility that the metallic alloy fuel on liquid phase formed by the metallurgical reactions can be fragmented without occurring the boiling of sodium on the surface of the melt.     

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.     

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.     

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.     

Precipitation Reaction of Uranyl Ion in Molten Potassium Thiocyanate

The qualitative and the quantitative analyses of the reaction products obtained by heating uranyl solution of molten potassium thiocyanate were carried out. It was found that in the presence of water in the melt, uranyl ion is converted into UO₂ accompanied with the evolution of CO₂ and the formation of free sulfur. The molar ratios on these products were almost equal to each other.

By detecting ammonia in reaction products, it was concluded that the precipitation reaction of uranyl ion in the melt is expressed as follows:

UO₂ ⁺⁺ + SCN^? + 2H₂O = UO₂ + CO₂ +S+ NH₄ ⁺.     

Phase segregation model and molten pool thermal-hydraulics during molten core-concrete interaction

A new model for melt thermal-hydraulics during molten core concrete interaction (MCCI) is presented. This model assumes that phase segregations occur in the melt, leading to a crust formation composed of refractory materials (UO₂ZrO₂). The interface temperature between this crust and the liquid melt is linked to the solid fraction and is calculated on the basis of a thermal equilibrium assumption. The solid fraction is also controlled by conduction heat transfer through the solid crust. It is shown that the temperatures measured in the ACE experiments are recalculated within a maximum deviation of 10%, (when referenced to the solidus temperature) without any adjustment. Other important consequences for this new approach are outlined: for physical properties, physico-chemical interactions, fission products behavior, mixing with sacrificial materials, and crust stabilities.     

Experiments on in-vessel melt coolability in the EC-FOREVER Program

This paper reports the results from the experiments conducted on the coolability of corium melt during a severe accident scenario when the bottom head is full of the core melt, undergoing natural circulation. These experiments are part of the EC-FOREVER Program in which vessel failure experiments have also been performed. The experiments are performed in a 1/10th scale vessel (400 mm diameter and 15 mm wall thickness) and the oxidic melt employed is the mixture CaO + B₂O₃ at 1400 K, representing the corium melt mixture of UO₂ + ZrO₂.The experiments employed an initial phase, during which uniform volumetric heating of the melt was provided and the vessel was pressurised to 25 bar, for several hours, to generate maximum creep deformation of 5%, in order to provide the conditions for the formation of a gap between the melt-pool crust and the bottom head wall. After this phase, the vessel was flooded with water.Data were obtained on only the vessel and the melt pool temperatures in one of the EC-FOREVER experiments reported here. In the second experiment, however, besides the temperature data, additional data were obtained on the steam flow rate and the heat transfer to the water, at the upper face of the melt pool, as a function of time.It was found that the gap cooling mechanism was not effective in reducing the vessel wall temperatures after water flooding. Post-test examinations revealed that the water ingression extended to the depth of only 60 mm in the melt pool. The character of the heat transfer to the water from the melt pool upper surface was found to be similar to that observed in the MACE tests for the coolability of an ex-vessel melt pool flooded by water at the top.     

Structural investigations of borosilicate glasses containing MoO3 by MAS NMR and Raman spectroscopies

High molybdenum concentration in glass compositions may lead to alkali and alkaline-earth molybdates crystallization during melt cooling that must be controlled particularly during the preparation of highly radioactive nuclear glassy waste forms. To understand the effect of molybdenum addition on the structure of a simplified nuclear glass and to know how composition changes can affect molybdates crystallization tendency, the structure of two glass series belonging to the SiO₂-B₂O₃-Na₂O-CaO-MoO₃ system was studied by ²⁹Si, ¹¹B, ²³Na MAS NMR and Raman spectroscopies by increasing MoO₃ or B₂O₃ concentrations. Increasing MoO₃ amount induced an increase of the silicate network reticulation but no significant effect was observed on the proportion of units and on the distribution of Na⁺ cations in glass structure. By increasing B₂O₃ concentration, a strong evolution of the distribution of Na⁺ cations was observed that could explain the evolution of the nature of molybdate crystals (CaMoO₄ or Na₂MoO₄) formed during melt cooling.     

Formation of diffusion barrier coating on superalloy 690 substrate and its stability in borosilicate melt at elevated temperature

Aluminized and thermally oxidized superalloy 690 substrates forming Al₂O₃ layer on (NiCr)Al + Cr₅Al₈ types aluminides and bare substrates were exposed in sodium borosilicate melt at 1248 K for 192 h. SEM–EDXS analysis along the cross-section of bare substrate with adhered glass revealed formation of a continuous, thick Cr₂O₃ layer at the substrate/glass interface due to its low solubility in borosilicate melt. XRD on aluminide coated and thermally oxidized specimen revealed existence of Al₂O₃ along with NiAl and Cr₅Al₈ type phases after the exposure in borosilicate melt. SEM–EDXS analysis along the cross-section of aluminide coated and thermally oxidized sample with adhered glass indicated good stability of coating in borosilicate melt without any phase formation at the coating/glass interface. However, some Al enrichment in glass phase adjacent to interface was noticed without any significant Ni or Cr enrichment.     

Gamma ray interaction with sodium phosphate glasses containing MoO3

Sodium metaphosphate glasses undoped and doped with varying MoO₃ contents were prepared and their UV-visible, infrared and Raman spectroscopy were measured. The ultraviolet-visible absorption were remeasured after each successive gamma irradiation. The induced color centers of the intrinsic defects of the undoped and extrinsic defects due to Mo-doped glasses were characterized. The structural forming units were derived and analyzed from infrared and Raman spectroscopic evolution. Also, some sodium phosphate glasses with varying Na₂O or P₂O₅ content and constant MoO₃ were investigated. The experimental results show that the parent undoped sodium metaphosphate reveals strong ultraviolet spectra originating from the presence of trace iron impurities. Upon gamma irradiation, some UV and visible induced absorption bands are resolved which can be correlated with POHC, OHC, PEC from phosphate network and extrinsic defect due to iron impurities. The high Mo-doped glass reveals two small visible bands with increasing the Na₂O content which seems to be related to polymolybdate ions. Infrared and Raman spectra show the characteristic absorption bands due mainly to metaphosphate network and molybdate groups.     

Investigation of the effect of B<Subscript>4</Subscript>C on the interaction of oxide melt and steel in a VVER-1000 vessel during a severe accident

The results of investigations of the interaction of U-Zr-B-C-O melts with steel, which are performed as part of the OECD Masca international program, are presented. It is found that, as a result of the interaction, boron and carbon become concentrated predominately in the metallic phase of the melt. As the initial mass ratio m_Fe/m_melt increases, the effect of the addition of B₄C on the melt-iron interaction decreases because the metallic phase is diluted with iron. It is concluded on the basis of a comparison of the results of the STFM-B Nos. 3, 7 experiments with the STFM-Fe Nos. 3, 7 experiments performed previously without the participation of boron carbide that the effect of boron carbide on the interaction of the oxide melt with iron decreases as the degree of oxidation of zirconium increases. Translated from Atomnaya énergiya, Vol. 105, No. 2, pp. 63–67, August, 2008.     

Natural convection heat transfer test for in-vessel retention at prototypic Rayleigh numbers – Results of COPRA experiments

Large-scale COPRA experiments were performed to investigate the natural convection heat transfer in melt pools for the in-vessel retention during severe accidents in Chinese large-scale advanced PWRs. Both water and binary mixture of 20 mol% NaNO₃ – 80 mol% KNO₃ were used as the melt simulant material in performed tests. Due to the full scale geometry of the COPRA test section, the Rayleigh numbers of the melt pool could reach up to the prototypic magnitude of 10¹⁶. Natural convection heat transfer tests at prototypic Rayleigh numbers have been performed to study the influence of the heat generation rate and melt simulant material on the melt pool temperature, heat flux distribution and heat transfer capability. The comparisons of the melt pool temperature and heat flux distribution from water experiments and molten binary salt experiments showed that the crust formation along the inner surface of the vessel wall could impact the heat transfer characteristics of the melt pool. And the heat flux distribution from COPRA water tests and molten salt tests were in good agreement with those from Jahn-Reineke water experiments and RASPLAV molten salt experiments, respectively. The heat transfer capability of the melt pool Nu_dn from COPRA molten salt tests were larger than those from water tests, but both were lower than those from ACOPO and BALI predictions within the same range of Rayleigh numbers (10¹⁵ – 10¹⁷).     

Zirconium carbide coating for corium experiments related to water-cooled and sodium-cooled reactors

Since the TMI and Chernobyl accidents the risk of nuclear severe accident is intensively studied for existing and future reactors. In case of a core melt-down accident in a nuclear reactor, a complex melt, called corium, forms. To be able to perform experiments with prototypic corium materials at high temperature, a coating which resists to different corium melts related to Generation I and II Water Reactors and Generation IV sodium fast reactor was researched in our experimental platforms both in IAE NNC in Kazakhstan and in CEA in France. Zirconium carbide was selected as protective coating for graphite crucibles used in our induction furnaces: VCG-135 and VITI. The method of coating application, called reactive wetting, was developed. Zirconium carbide revealed to resist well to the (U_x, Zr_y)O_2−z water reactor corium. It has also the advantage not to bring new elements to this chemical system. The coating was then tested with sodium fast reactor corium melts containing steel or absorbers. Undesirable interactions were observed between the coating and these materials, leading to the carburization of the corium ingots. Concerning the resistance of the coating to oxide melts without ZrO₂, the zirconium carbide coating keeps its role of protective barrier with UO₂-Al₂O₃ below 2000 °C but does not resist to a UO₂-Eu₂O₃ mixture.     

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.     

Interaction of a model melt of a reactor core with zirconium dioxide ceramic with different porosity

The results of an experimental investigation of the interaction of zirconium dioxide ceramic with different porosity with a model composition of a core melt are presented. The experiments were performed with melt composition (in mass%) UO₂ 46.6, ZrO₂17.6, and Fe₂O₃ under isothermal conditions at 1800°C in an argon atmosphere. Data were obtained on the rate of erosion of the dense ceramic, the character of the permeation of the pores and the pore morphology, and the distribution of the melt elements along the height of the porous layer, 11 figures, 1 table, 3 references. Deceased. Russian Science Center “Kurchatov Institute.” Translated from Atomnaya énergiya, Vol. 88, No. 4, pp. 266–277, April, 2000.