Analysis of molten pool physico-chemical interactions and interpretation of the Phebus FP tests observations

The new 2D model for non-equilibrium U–Zr–O melt oxidation is developed on the base of the previously developed 1D model and extended to consideration of the general case of simultaneous UO₂ dissolution and melt oxidation accompanied with growth of the peripheral ceramic layer (crust) and bulk ceramic precipitates. The model is validated against isothermal crucible tests where precipitation of ceramic phase during molten Zr interactions with ceramic crucible walls along with oxide growth on non-oxidised surface of the melt were observed. In order to analyse melt oxidation under transient temperature conditions of the Phebus FP tests, the new physico-chemical model was tightly coupled with the heat exchange model and applied to interpretation of the post-test microstructure observations of the molten pool in the FP bundles.     

Corium phase equilibria based on MASCA, METCOR and CORPHAD results

Experimental data on component partitioning between suboxidized corium melt and steel in the in-vessel melt retention (IVR) conditions are compared. The data are produced within the OECD MASCA program and the ISTC CORPHAD project under close-to-isothermal conditions and in the ISTC METCOR project under thermal gradient conditions. Chemical equilibrium in the U–Zr–Fe(Cr,Ni,…)–O system is reached in all experiments. In MASCA tests the molten pool formed under inert atmosphere has two immiscible liquids, oxygen-enriched (oxidic) and oxygen-depleted (metallic), resulting of the miscibility gap of the mentioned system. Sub-system data of the U–Zr–Fe(Cr,Ni,…)–O phase diagram investigated within the ISTC CORPHAD project are interpreted in relation with the MASCA results. In METCOR tests the equilibrium is established between oxidic liquid and mushy metallic part of the system. Results of comparison are discussed and the implications for IVR noted.     

U–Zr–Fe–O Melt density

A method of evaluating the density of U–Zr–Fe–O melts, which is of interest for predicting the spatial configuration of stratified melt during a serious accident in water moderated and cooled reactors, has been proposed previously. In the present work, the density of Fe–U–Zr–O metallic melt at 2023 K, (U, Zr)O₂ melt with atomic ratio U/Zr = 0.9 at 2973 K, and zirconium oxide melt at 3073 K are determined experimentally in order to verify the proposed model. The agreement between the calculations and the experimental results is satisfactory as a whole; this makes it possible to recommend the proposed method for evaluating the spatial configuration of the melt in a VVER vessel during a serious accident.     

Core loss during a severe accident (COLOSS)

The COLOSS project was a 3-year shared-cost action, which started in February 2000. The work-programme performed by 19 partners was shaped around complementary activities aimed at improving severe accident codes. Unresolved risk-relevant issues regarding H₂ production, melt generation and the source term were studied through a large number of experiments such as (a) dissolution of fresh and high burn-up UO₂ and MOX by molten Zircaloy, (b) simultaneous dissolution of UO₂ and ZrO₂, (c) oxidation of U–O–Zr mixtures, (d) degradation–oxidation of B₄C control rods.Corresponding models were developed and implemented in severe accident computer codes. Upgraded codes were then used to apply results in plant calculations and evaluate their consequences on key severe accident sequences in different plants involving B₄C control rods and in the TMI-2 accident.Significant results have been produced from separate-effects, semi-global and large-scale tests on COLOSS topics enabling the development and validation of models and the improvement of some severe accident codes. Break-throughs were achieved on some issues for which more data are needed for consolidation of the modelling in particular on burn-up effects on UO₂ and MOX dissolution and oxidation of U–O–Zr and B₄C–metal mixtures. There was experimental evidence that the oxidation of these mixtures can contribute significantly to the large H₂ production observed during the reflooding of degraded cores under severe accident conditions.The plant calculation activity enabled (a) the assessment of codes to calculate core degradation with the identification of main uncertainties and needs for short-term developments and (b) the identification of safety implications of new results.Main results and recommendations for future R&D activities are summarized in this paper.     

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.     

Model for melt blockage (slug) relocation and physico-chemical interactions during core degradation under severe accident conditions

The model describing massive melt blockage (slug) relocation and physico-chemical interactions with steam and surrounding fuel rods of a bundle is developed on the base of the observations in the CORA tests. Mass exchange owing to slug oxidation and fuel rods dissolution is described by the previously developed 2D model for the molten pool oxidation. Heat fluxes in oxidising melt along with the oxidation heat effect at the melt relocation front are counterbalanced by the heat losses in the surrounding media and the fusion heat effect of the Zr claddings attacked by the melt. As a result, the slug relocation velocity is calculated from the heat flux matches at the melt propagation front (Stefan problem). A numerical module simulating the slug behaviour is developed by tight coupling of the heat and mass exchange modules. The new model demonstrates a reasonable capability to simulate the main features of the massive slug behaviour observed in the CORA-W1 test.     

Influence of corium oxidation on fission product release from molten pool

Qualitative and quantitative determination of the release of low-volatile fission products and core materials from molten oxidic corium was investigated in the EVAN project under the auspices of ISTC. The experiments carried out in a cold crucible with induction heating and RASPLAV test facility are described. The results are discussed in terms of reactor application; in particular, pool configuration, melt oxidation kinetics, critical influence of melt surface temperature and oxidation index on the fission product release rate, aerosol particle composition and size distribution. The relevance of measured high release of Sr from the molten pool for the reactor application is highlighted. Comparisons of the experimental data with those from the COLIMA CA-U3 test and the VERCORS tests, as well as with predictions from IVTANTHERMO and GEMINI/NUCLEA codes are made. Recommendations for further investigations are proposed following the major observations and discussions.     

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.     

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.     

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.     

Plasma–Hydrogen Reduction of Waste Uranium Hexafluoride

A technological scheme for plasma–hydrogen reduction of ²³⁵U waste uranium hexafluoride, producing metallic uranium and water-free hydrogen fluoride, is proposed. The results of experimental investigations of the basic stages of this technological scheme are examined: production of U–F–H plasma flow and production and separation of uranium melt and water-free hydrogen fluoride. The level of plasma and high-frequency technology for implementing the plasma–hydrogen technology for converting waste UF₆ into metallic uranium and water-free HF is analyzed.     

Numerical analysis of fragmentation mechanisms in vapor explosions

Fragmentation of molten metal is the key process in vapor explosions. However, this process is so rapid that the mechanisms have not yet been clarified in experimental studies. In addition, numerical simulation is difficult because we have to analyze water, steam and molten metal simultaneously with boiling and fragmentation. The authors have been developing a new numerical method, the moving particle semi-implicit (MPS) method, based on moving particles and their interactions. Grids are not necessary. Incompressible flows with fragmentation on free surfaces have been calculated successfully using the MPS method. In the present study, numerical simulation of the fragmentation processes using the MPS method is carried out to investigate the mechanisms. A numerical model to calculate boiling from water to steam is developed. In this model, new particles are generated on water–steam interfaces. A two-step pressure calculation algorithm is also developed. Pressure fields are separately calculated in both heavy and light fluids to maintain numerical stability with the water and steam system. The new model and algorithm are added to the MPS code. Water jet impingement on a molten tin pool is calculated using the MPS code as a simulation of collapse of a vapor film around a melt drop. Penetration of the water jet, which is assumed in Kim–Corradini’s model, is not observed. If the jet fluid density is hypothetically larger, the penetration appears. Next, impingement of two water jets is calculated. A filament of the molten metal is observed between the two water jets as assumed in Ciccarelli–Frost’s model. If the water density is hypothetically larger, the filament does not appear. The critical value of the density ratio of the jet fluid over the pool fluid is ρ_jet/ρ_pool=0.7 in this study. The density ratios of tin–water and UO₂–water are in the region of filament generation, Ciccarelli–Frost’s model. The effect of boiling is also investigated. Growth of the filament is not accelerated when the normal boiling is considered. This is because normal boiling requires more time than that of the jet impingement, although the filament growth is governed by an instant of the jet impingement. Next, rapid boiling based on spontaneous nucleation is considered. The filament growth is markedly accelerated. This result is consistent with the experimental fact that the spontaneous nucleation temperature is a necessary condition of vapor explosions.     

Synopsis and outcome of the QUENCH experimental program

The paper gives an overview of the main outcome of the QUENCH program launched in 1997 at the Karlsruhe Institute of Technology (KIT), formerly Karlsruhe Research Center (FZK). The research program comprises bundle experiments as well as complementary separate-effects tests. The focus of the experiments performed from 1997 to 2009 was on scenarios of severe accidents whereas that of the future test program will be on large-break loss-of-coolant accidents (LOCA) in the frame of design-basis accidents, and debris coolability, in the frame of severe accidents. The major objective of the program is to deliver experimental and analytical data to support the development and validation of quench and quench-related models as used in code systems that model severe accident progression in light water reactors.So far, 15 integral bundle QUENCH experiments with 21-31 electrically heated fuel rod simulators of 2.5 m length have been conducted. The following parameters and their influence on bundle degradation and reflood have been investigated: degree of pre-oxidation, temperature at initiation of reflood, flooding rate, influence of neutron absorber materials (B₄C, AgInCd), air ingress, and influence of the type of cladding alloy.In six tests, reflooding of the bundle led to a temporary temperature excursion driven by runaway oxidation of zirconium alloy components and resulting in release of a significant amount of hydrogen, typically two orders of magnitude greater than in those tests with “successful” quenching in which cool-down was rapidly achieved. Considerable formation, relocation, and oxidation of melt were observed in all tests with escalation. The temperature boundary between rapid cool-down and temperature escalation was typically in the range of 2100-2200 K in the “normal” quench tests, i.e. in tests without absorber and/or steam starvation. Tests with absorber and/or steam starvation were found to lead to temperature escalations at lower temperatures.All phenomena occurring in the bundle tests have been investigated additionally in parametric and more systematic separate-effects tests. Oxidation kinetics of various cladding alloys, including advanced ones, have been determined over a wide temperature range (873-1773 K) in different atmospheres (steam, oxygen, air, and their mixtures). Hydrogen absorption by different zirconium alloys was investigated in detail, recently also using neutron radiography as non-destructive method for determination of hydrogen distribution in claddings. Furthermore, degradation mechanisms of absorber rods including B₄C and AgInCd as well as the oxidation of the resulting low-temperature melts have been studied. Steam starvation was found to cause deterioration of the protective oxide scale by thinning and chemical reduction.The most recent topic of the QUENCH program has been investigation of the behavior of advanced cladding materials (ACM) in comparison with the classical Zircaloy-4. Although separate-effects tests have shown some differences in oxidation kinetics, the influence of the various cladding alloys on the integral bundle behavior during oxidation and reflooding was only limited.     

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.     

On the formation of U–Al alloys in the molten LiCl–KCl eutectic

U–Al alloy formation has been studied in the temperature range of 400–550 °C by electrochemical techniques in the molten LiCl–KCl eutectic. Cyclic voltammetry showed that underpotential reduction of U(III) onto solid Al occurs at a potential about 0.35 V more anodic than pure U deposition. Open circuit potential measurements, recorded after small depositions of U metal onto the Al electrode, did not allow the distinction between potentials associated with UAl_x alloys and the Al rest potential, as they were found to be practically identical. As a consequence, a spontaneous chemical reaction between dissolved UCl₃ and Al is thermodynamically possible and was experimentally observed. Galvanostatic electrolyses were carried out both on Al rods and Al plates. Stable and dense U–Al deposits were obtained with high faradic yields, and the possibility to load the whole bulk of a thin Al plate was demonstrated. The analyses (by SEM-EDX and XRD) of the deposits indicated the formation of different intermetallic phases (UAl₂, UAl₃ and UAl₄) depending on the experimental conditions.     

BETA experiments on melt-concrete interaction: the role of zirconium and the potential sump water contact during core melt-down accidents

In the BETA test facility of Kernforschungszentrum Karlsruhe, prototypical core melts can be simulated in concrete structures sufficient in size to allow a computer-code-assisted extrapolation to be made to the reactor geometry.Three experiments have been carried out to investigate special aspects of molten corium interacting with concrete. The investigations and measurements show the dominance of Zr oxidation during concrete attack by the chemical reduction of SiO₂ to elemental Si and the subsequent Si oxidation by the gases from the concrete.Additionally, the failure of a cylindrical concrete wall was studied, which is eroded on the inner side by a heated melt while being cooled outside by stagnant water. In the experiment wall failure occurs and the melt relocates into the water annulus.Application of the experimental results to light-water reactor severe accidents is discussed.     

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.     

Thermal compatibility of centrifugally atomized U–Mo powders with aluminium in a dispersion fuel

The thermal compatibility of centrifugally atomized U–Mo alloys with aluminum has been studied. Samples of extruded dispersions of 24 vol.% spherical U–2 wt.% Mo and U–10 wt.% Mo powders in an aluminum matrix were annealed for over 2000 h at 400°C. No significant dimensional changes occurred in the U–10 wt.% Mo/aluminum dispersions. The U–2 wt.% Mo/aluminum dispersion, however, increased in volume by 26% after 2000 h at 400°C. This large volume change is mainly due to the formation of voids and cracks resulting from nearly complete interdiffusion of U–Mo and aluminum. Interdiffusion between U–10 wt.% Mo and aluminum was found to be minimal. The different diffusion behavior is primarily due to the fact that U–2 wt.% Mo decomposes from an as-atomized metastable γ-phase (bcc) solid solution into the equilibrium α-U and U₂Mo two-phase structure during the experiment, whereas U–10 wt.% Mo retains the metastable γ-phase structure throughout the 2000 h anneal and thereby displays superior thermal compatibility with aluminum compared to U–2 wt.% Mo.     

Experimental results of integral effects tests with 1/10th scale zion subcompartment structures in the Surtsey test facility

Three integral effects tests (IET-1, IET-3, and IET-6) were conducted to investigate the effects of high-pressure melt ejection on direct containment heating. A 1:10 linear scale model of the Zion reactor pressure vessel (RPV), cavity, instrument tunnel, and subcompartment structures were constructed in the Surtsey test facility at Sandia National Laboratories. The RPV was modeled with a melt generator that consisted of a steel pressure barrier, a cast MgO crucible, and a thin steel inner liner. The melt generator/crucible had a hemispherical bottom head containing a graphite limitor plate with a 4 cm exit hole to simulate the ablated hole in the RPV bottom head that would be formed by tube ejection in a severe nuclear power plant accident. The reactor cavity model contained 3.48 kg water with a depth of 0.9 cm that corresponded to condensate levels in the Zion plant. 43 kg iron oxide/aluminum/ chromium thermite was used to simulate molten core debris. The molten thermite in the three tests was driven into the scaled reactor cavity by slightly superheated steam at 7.1, 6.1, and 6.3 MPa for IET-1, IET-3, and IET-6 respectively. The IET-1 atmosphere was pre-inerted with nitrogen, while the IET-3 atmosphere was nitrogen with approximately 9.0 mol% O₂. The IET-6 atmosphere was nitrogen with 9.79 mol% O₂ and 2.59 mol% pre-existing hydrogen. In IET-1, approximately 233 g mol hydrogen were produced but almost none burned because oxygen was not available. In IET-3, approximately 227 g mol hydrogen were produced and 190 g mol burned. In IET-6, approximately 319 g mol hydrogen were produced and 345 g mol burned. The peak pressure increases in the IET-1, IET-3 and IET-6 experiments were 0.098, 0.246, and 0.279 MPa respectively. In IET-3 and IET-6 hydrogen burned as it was pushed out of the subcompartments into the upper region of the Surtsey vessel. In IET-6, although a substantial amount of pre-existing hydrogen burned, it apparently did not burn on a time scale that made a significant contribution to the peak pressure increase in the vessel.     

The feasibility of small size specimens for testing of environmentally assisted cracking of irradiated materials and of materials under irradiation in reactor core

Environmentally assisted cracking (EAC) or, in other words, stress corrosion cracking (SCC) of in-core materials has become an increasingly important reason for the downtime and maintenance costs of nuclear power plants (NPPs). Use of small size specimens for stress corrosion testing of irradiated materials is necessary because handling of high dose rate materials is difficult and the availability of these materials is limited. A drawback of using small size specimens is that they do not in some cases fulfil the requirements of the relevant testing standards and sometimes their limited load-bearing capacity prevents corrosion fatigue tests and tests with static loading at reasonable K_I values. The test results show that the ductile fracture resistance curves of a Cu–Zr–Cr alloy are, to some extent, independent of the specimen geometry and size. However, the curves of small specimens deviate from the curves of larger specimens at high J values (large plastic zone relative to the remaining ligament) or when the crack growth exceeds about 30% of the remaining ligament. The size dependency of the tested Cu–Zr–Cr alloy seems to be a consequence of decreasing stress triaxiality as the size of the specimen is decreased. The results of the SCC tests of sensitized SIS 2333 stainless steel (equal to AISI 304) specimens in simulated boiling water reactor (BWR) water show that the plastic deformation of the remaining ligament of the specimen has no significant effect on the environmentally assisted crack growth rate. This indicates that stress corrosion testing is not limited by the specimen size. The size dependency in SCC tests should be further studied by conducting tests using various specimen sizes.