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.     

Heat flux capabilities of first-wall tube arrays for an experimental fusion reactor

A preliminary design study has been made of some of the thermomechanical problems of water and helium cooling for the first wall of a near-term experimental fusion reactor. The first wall is envisioned as an array of 316 stainless steel tubes between the plasma and the blanket modules to intercept a heat flux from the plasma estimated to be between 0.25 and 1.0 MW/m². Evaluations have been made of the maximum allowable heat fluxes for constraints imposed on the tube wall temperature, the cyclic stresses, the quasi-steady stresses and energy recovery from the coolant. For tubes with 2 meter long heated sections, 10 mm inside diameter and 1 mm wall thickness, water cooling was found to be more than adequate for plasma heat fluxes over 1 MW/m² with a fatigue life of 10⁶ cycles; for a 2 mm wall thickness, at least 0.7 MW/m² can be handled for the same fatigue life. Helium-cooled tubes can also handle heat fluxes up to about 1 MW/m² with a 1 mm tube wall thickness and over 0.5 MW/m² with a 2 mm tube wall thickness, but the required pumping powers tend to be high. The problems of plasma disruptions and erosion by energetic plasma ions are also discussed briefly.     

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).     

Void fraction and heat transport in two-dimensional mixed size particle beds with internal heat sources

The paper discusses the boiling heat transfer from a porous bed with internal heat sources and refers to the configuration in a nuclear reactor after a partial core melt. The flow of coolant, the temperature and the local liquid/vapor distribution were investigated in a two-dimensional configuration. Experiments were conducted using monodisperse beds as well as a mixture of two different particle sizes with a total porosity below 20%. In some tests the bed was supported by a shell of porous material to create a gap along the bottom of the test container. Water was used for tests up to 9% of the critical pressure, while other tests were made with R134a up to 44% of the critical pressure. The maximum heating rate realized inductively was 730 kW/m². The experiments have been compared to analytical results with a one-dimensional approach.It is shown that in contrary to the situation in small cylindrical configurations the heat transfer was increased by large buoyancy driven convective flows. If there was a gap along the container bottom an additional flow of liquid improved the coolability of the bottom region even if the upper part of the particle bed was already overheated. In case of high density ratios (water at low pressure), the measurements indicated a strong enhancement of the coolant flow above a certain minimum heating rate resulting in decreasing vapor fraction values which were nearly independent of the system pressure. This was assumed to be caused by the appearance of vertical channels through which the vapor could flow through the particle bed.     

Heat transport and void fraction in granulated debris

The paper discusses the boiling heat transfer from a porous bed with internal heat sources and refers to the configuration in a nuclear reactor after a partial core melt. The flow of coolant, the temperature and the local liquid/vapor distribution were investigated in a two-dimensional configuration. Experiments were conducted using monodisperse beds as well as a mixture of two different particle sizes with a total porosity below 20%. In some tests the bed was supported by a shell of porous material to create a gap along the bottom of the test container. Water was used for tests up to 9% of the critical pressure, while other tests were made with R134a up to 44% of the critical pressure. The maximum heating rate realized inductively was 730 kW/m². The experiments have been compared to analytical results with a one-dimensional approach.It is shown that in contrary to the situation in small cylindrical configurations the heat transfer was increased by large buoyancy driven convective flows. If there was a gap along the container bottom an additional flow of liquid improved the coolability of the bottom region even if the upper part of the particle bed was already overheated. In case of high density ratios (water at low pressure), the measurements indicated a strong enhancement of the coolant flow above a certain minimum heating rate resulting in decreasing vapor fraction values which were nearly independent of the system pressure. This was assumed to be caused by the appearance of vertical channels through which the vapor could flow through the particle bed.     

An experimental study on pressure drop and dryout heat flux of two-phase flow in packed beds of multi-sized and irregular particles

This paper is concerned with debris bed coolability in a postulated severe accident of light water reactors, where the debris particles are irregular and multi-sized. To obtain and verify the friction laws predicting the hydrodynamics of the debris beds, the drag characteristics of air/water single- and two-phase flow in a particulate bed packed with multi-sized spheres or irregular sand particles were investigated on the POMECO-FL test facility. The same types of particles were then loaded in the test section of the POMECO-HT facility to obtain the dryout heat fluxes of the particulate beds heated volumetrically. The effective (mean) particle diameter is 2.25 mm for the multi-sized spheres and 1.75 mm for the sand particles, determined from the Ergun equation and the measured pressure drop of single-phase flow through the packed bed. Given the effective particle diameter, both the pressure drop and the dryout heat flux of two-phase flow through the bed can be predicted by the Reed model. The experiment also shows that the bottom injection of coolant improves the dryout heat flux significantly and the first dryout position is moving upward with increasing bottom injection flowrate. Compared with top-flooding case, the dryout heat flux of the bed can be doubled if the superficial velocity of coolant injection is 0.21–0.27 mm/s. The experimental data provides insights for interpretation of debris bed coolability (how to deal with the multi-sized irregular particles), as well as high-quality data for validation of the coolability analysis models and codes.     

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 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.     

Experimental and Analytical Investigation of Formation and Cooling Phenomena in High Temperature Debris Bed

ABSTRACT

Key phenomena in the cooling states of underwater debris beds were classified based on the premise that a target debris bed has a complicated geometry, nonhomogeneous porosity, and volumetric heat. These configurations may change due to the molten jet breakup, droplet agglomeration, anisotropic melt spreading, two-phase flow in a debris bed, particle self-leveling and penetration of molten metals into a particle bed. Based on these classifications, the modular code system THERMOS was designed for evaluating the cooling states of underwater debris beds. Three tests, DEFOR-A, PULiMS, and REMCOD were carried in six phases to extend the existing database for validating implemented models. Up to Phase-5, the main part of these tests has been completed and the test plan has been modified from the original one due to occurrences of unforeseeable phenomena and changes in test procedures. This paper summarizes the entire test plan and representative data trends prior to starting individual data analyses and validations of specific models that are planned to be performed in the later phases. Also, it tries to timely report research questions to be answered in future works, such as various scales of melt-coolant interactions observed in the shallow pool PULiMS tests.     

CHF enhancement by honeycomb porous plate in saturated pool boiling of nanofluid

One strategy for severe accidents is in-vessel retention (IVR) of corium debris. In order to enhance the capability of IVR in the case of a severe accident involving a light-water reactor, methods to increase the critical heat flux (CHF) should be considered. Approaches for increasing the IVR capability must be simple and installable at low cost. Moreover, cooling techniques for IVR should be applicable to a large heated surface. Therefore, as a suitable cooling technology for required conditions, we proposed cooling approaches using a honeycomb porous plate for the CHF enhancement of a large heated surface in a saturated pool boiling of pure water. In this paper, CHF enhancement by the attachment of a honeycomb-structured porous plate to a heated surface in saturated pool boiling of a TiO₂-water nanofluid was investigated experimentally under atmospheric pressure. As a result, the CHF with a honeycomb porous plate increases as the nanoparticle concentration increases. The CHF is enhanced significantly up to 3.2 MW/m² at maximum upon the attachment of a honeycomb porous plate with 0.1 vol.% nanofluid. To the best of the author's knowledge, under atmospheric pressure, a CHF of 3.2 MW/m² is the highest value for a relatively large heated surface having a diameter exceeding 30 mm.     

A study on concrete degradation during molten core/concrete interactions

A series of MCCI tests was performed in COTELS project at NUPEC to examine concrete degradation characteristics during MCCI with and without water addition onto the debris. Molten stainless steel or a mixture composed of UO₂, ZrO₂, Zr and stainless steel was slumped into a two-dimensional concrete trap, where volumetric decay heat generation was simulated by an induction heating technique. The results of dry MCCI tests implied that concrete ablation was dominated by melting of aggregates when the debris was crusted and cement was thermally weaker than aggregates. Without presence of stable crust, unmolten aggregates were possible to relocate upward due to the density difference from the debris. Concrete responses under a wet condition showed a tendency that water migrated into thermally degraded concrete. A preliminary water migration model was incorporated into COCO code for transient heat conduction. The prediction by COCO code agreed with the tendency of concrete thermal responses observed in the dry and wet MCCI tests.     

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.     

Design and evaluation of an optimized W/Cu interlayer for W monoblock components

Divertor plasma-facing components of future fusion reactors should be able to withstand heat fluxes of 10-20 MW/m² in stationary operation. Tungsten blocks with an inner cooling tube made of CuCr1Zr, so-called monoblocks, are potential candidates for such water-cooled components. To increase the strength and reliability of the interface between the W and the cooling tube of a Cu-based alloy (CuCr1Zr), a novel advanced W-fibre/Cu metal matrix composite (MMC) was developed for operation temperatures up to 550 °C. Based on optimization results to enhance the adhesion between fibre and matrix, W fibres (W_f) were chemically etched, coated by physical vapour deposition with a continuously graded W/Cu_PVD interlayer and then heated to 800 °C. The W_f/Cu MMC was implemented by hot-isostatic pressing and brazing process in monoblock mock-ups reinforcing the interface between the plasma-facing material and the cooling channel. The suitability of the MMC as an efficient heat sink interface for water-cooled divertor components was tested in the high heat flux (HHF) facility GLADIS. Predictions from finite element simulations of the thermal behaviour of the component under loading conditions were confirmed by the HHF tests. The W_f/Cu MMC interlayer of the mock-ups survived cyclic heat loads above 10 MW/m² without any damage. One W block of each tested mock-up showed stable thermal behaviour at heat fluxes of up to 10.5 MW/m².     

Dryout heat flux experiments with deep heterogeneous particle bed

A test facility has been constructed at Technical Research Centre of Finland (VTT) to simulate as accurately as possible the ex-vessel core particle bed in the conditions of Olkiluoto nuclear power plant. The STYX particle bed reproduces the anticipated depth of the bed and the size range of particles having irregular shape. The bed is immersed in water, creating top flooding conditions, and internally heated by an array of electrical resistance heating elements. Dryout tests have been successfully conducted at 0.1–0.7 MPa pressure for both uniformly mixed and stratified bed geometries. In all tests, including the stratified ones, the dry zone first formed near the bottom of the bed. The measured dryout heat fluxes increased with increasing pressure, from 232 kW/m² at near atmospheric pressure to 451 kW/m² at 0.7 MPa pressure. The data show some scatter even for the uniform bed. The tests with the stratified bed indicate a clear reduction of critical power due to the presence of a layer of small particles on top of the uniform bed. Comparison of data with various critical power (dryout heat flux) correlations for porous media shows that the most important parameter in the models is the effective particle diameter. Adiabatic debris bed flow resistance measurements were conducted to determine the most representative particle diameter. This diameter is close, but not equal, to the particle number-weighted average diameter of the bed material. With it, uniform bed data can be calculated to within an accuracy of 3–28% using Lipinski's 0-D model. In the stratified bed experiments, it appears that the top layer was partially fluidized, hence the measured critical power was significantly higher than calculated. Future experiments are being planned with denser top layer material to eliminate non-prototypic fluidization.     

Measurements of solid particle suspension concentrations in boiling pools

An experimental study was performed to measure vertical particle suspension concentrations in contaminated boiling pools. The study was conducted in a nucleate boiling regime. Tests were conducted with distilled water and solid nickel particles ranging in size from 5 to 40 μm. The test apparatus was 30 cm wide and 120 cm high. Particle concentrations were on the order of milligrams per cubic centimeter of slurry. Heat fluxes on the order of 100 kW m⁻² were attained by electrical bottom heating. Corresponding superficial vapor velocities were on the order of several centimeters per second. Measurements were aimed to reveal the dependence of dilute particle suspension distributions on the pool depth, heat flux and particle loading. The results indicate that only a fraction of the particle loading gets suspended. This fraction turns out to be largely dependent on the pool depth and heat flux, but is insensitive to particle loading variations. The particle distribution within the suspension was analyzed with the dispersion–sedimentation model. The work demonstrates the applicability of this model for the conditions of boiling pools and partial particle suspensions. Furthermore, the results indicate that for shallow pools and small size particles the distributions are fairly uniform. Consequently, the suspension concentration is substantially a function of the fraction suspended. The work provides experimental data to evaluate this fraction for a range of heat fluxes and pool depths. The knowledge of particle suspension concentrations has important industrial and environmental applications in the power generation, nuclear and chemical engineering industries. For example, the results could be used to evaluate the amount of particle released from a contaminated boiling-pool spill, or could be used to reduce conservatism applied in analyses of pools containing larger particles.     

Medium-scale experiment with production of prototypical ~100 kg melt and an analysis of its interaction with concrete

The results of a medium-scale experiment in which a prototypical melt is produced by combustion of a chemically active substance in the course of the reaction 2Fe₂O₃ + 3Zr = 3ZrO₂ + 4Fe + 2840 kJ/kg in a ~6·10⁻² m³ concrete container are presented. It is shown that the ~100 kg melt, whose temperature is 2700–3200 K, so obtained produces heat fluxes 100–150 kW/m2 into the walls and bottom of the concrete container for ~10 min. The ablation of the walls of the concrete container was 2.5–3 cm at the completion of the experiment.     

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.     

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.     

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.