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%.     

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.     

Simulation of MCCI with the TOLBIAC-ICB code based on the phase segregation model

In the event of a severe accident in a pressurized water reactor, corium, a mixture of molten materials issued from the fuel, cladding and structural elements, appears in the reactor core. In some circumstances, corium is likely to melt through the reactor pressure vessel and spread over the concrete basemat of the reactor pit. Molten core concrete interaction (MCCI) then occurs. The main question that has to be addressed in this scenario is whether and when the corium will make its way through the basemat. For some years, CEA is developing a numerical code named TOLBIAC-ICB in order to simulate molten core concrete interaction in reactor case. The general approach used in this code is based on the phase segregation model developed by CEA. The solid phase is supposed to be located at the corium pool boundaries as a solid crust composed of refractory oxides, whereas the corium pool contains no solid. The interfacial temperature between the crust and the pool is the liquidus temperature calculated with the composition of the pool. The interaction between thermalhydraulics (mass and energy balances) and physico-chemistry (liquidus temperature, crust composition, chemical reaction) is modelled through a coupling between TOLBIAC-ICB and the GEMINI code for the determination of the physico-chemistry variables. The main purpose of this paper is to present the modelling used in TOLBIAC-ICB and some validation calculations using the data of experiments available in the literature.     

Ex-vessel core melt stabilization research (ECOSTAR)

The project on ex-vessel core melt stabilization research (ECOSTAR) started in January 2000 to be concluded by end of 2003. The project is performed by 14 partner institutions from five European countries and involves a large number of experiments with low- and high-temperature simulant melts and real corium at different scales. Model development and scaling analysis allows application of the research results to existing and to future LWRs in the area of reactor design and accident mitigation. The project is oriented toward the analysis and mitigation of severe accident sequences that could occur in the ex-vessel phase of a postulated core melt accident. The issues are: (1) the release of melt form the pressure vessel, (2) the transfer and spreading of the melt on the basement, (3) the analysis of the physical–chemical processes that are important for corium behavior especially during concrete erosion with onset of solidification, and (4) stabilization of the melt by cooling through direct water contact. The results achieved so far resolve a number of important issues: the amount of melt that could be transferred at RPV failure from the RPV into the containment can be substantially reduced by lowering the residual pressure in the primary circuit. It is found that melt dispersion also strongly depends on the location of the RPV failure, and that lateral failure results in substantially less melt dispersion. During melt release, the impinging melt jet could erode parts of the upper basement surface. Jet experiments and a derived heat transfer relation allow estimation of its contribution to concrete erosion. Spreading of the corium melt on the available basement surface is an important process, which defines the initial conditions for concrete attack or for the efficiency of cooling in case of water contact, respectively. Validation of the spreading codes based on a large-scale benchmark experiment is underway and will allow determination of the initial conditions, for which a corium melt can be assumed to spread homogeneously over the available surface. Experiments with UO₂-based corium melts highlight the role of phase segregation during onset of melt solidification and during concrete erosion. To cool the spread corium melt, the efficacy of top flooding and bottom flooding is investigated in small-scale and in large-scale experiments, supported by model developments. Project assessment is continuing to apply the results to present and future reactors.     

Controlling and cooling core melts outside the pressure vessel

For future reactors, the control and cooling of ex-vessel corium melts is under consideration to increase the passive safety features even for very unlikely severe accidents. In this context, different research activities are studying ex-vessel corium behaviour and control, including the implementation of a core cooling device outside the reactor pressure vessel in order to prevent basement erosion and to maintain the integrity of the containment. This paper describes current research on key phenomena which must be understood and quantified to be finally controlled by the cooling device. These are the release of corium melt from the pressure vessel, the temporary retention of the melt in the reactor cavity until melt through of the gate, spreading of the melt on a large surface, and finally the cooling and solidification of the melt by direct water contact. The experiments use high temperature melts which are similar to corium melts. Where necessary, models are developed to transfer the results to reactor scale.     

An analysis of containment responses during a station blackout accident

An analysis of the responses of the containment during a station blackout accident is performed for the APR1400 nuclear power plant using MELCOR 2.1. The analysis results show that the containment failure occurs at about 84.14 h. Prior to the failure of the reactor vessel, the containment pressure increases slowly. Then, a rapid increase of the containment pressure occurs when a large amount of hot molten corium is discharged from the reactor pressure vessel to the cavity. The molten corium concrete interaction (MCCI) is arrested when water is flooded over a molten corium in the cavity. The boiling of water in the cavity causes a fast increase in the containment pressure. During the early phase of the accident, a large amount of steam is condensed inside the containment due to the presence of the heat structures. This results in a mitigation of a containment pressure increase. During the late phase, the containment pressure increases gradually due to the addition of steam and gases from an MCCI and water evaporation. It was found that two-thirds of the total mass of steam and gases in the containment is from an MCCI and one-third of the mass is from water evaporation.     

Development and verification of molten corium–concrete interaction code

During a severe accident of Pressurized Water Reactor(PWR), the core materials was heated, melt located on the lower head of Reactor Pressure Vessel(RPV). With the temperature rise, the corium will melt through the lower head and discharge into the reactor cavity. Those corium will interact with the concrete and damage the integrity of the containment, so some coolability method should used to quench the corium. In order to investigate the progress of MCCI, a MCCI analysis code, that is MOCO, was developed. The MOCO includes the heat transfer behavior in axial and radial directions from the molten corium to the basemat and sidewall concrete, crust generation and growth, and coolability mechanisms reveal the concrete erosion and gas release, which are important for the interaction process. Cavity ablation depth, melt temperature, and gas release are the key parameters in the interaction research. The physical-chemistry reaction is also involved in MOCO code. In the present paper, the related MCCI experiment data were used to verify the models of the MOCO and the calculation results of MOCO were also compared with other MCCI analysis codes.     

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.     

The COMET-L3 experiment on long-term melt-concrete interaction and cooling by surface flooding

The COMET-L3 experiment considers the long-term situation of corium/concrete interaction in an anticipated core melt accident of a light water reactor after the metal melt is layered beneath the oxide melt. The experimental focus is on the cavity formation in the basemat and the risk of a long-term basemat penetration by the metallic part of the melt. The experiment investigates the two-dimensional concrete erosion in a cylindrical crucible of 60 cm in diameter fabricated from siliceous concrete in the first phase of the test, and the influence of surface flooding in the second phase. The initial mass of the melt was 425 kg steel and 211 kg oxide. Decay heating in the two-component metal and oxide melt is simulated by sustained induction heating of the metal phase that is overlaid by the oxide melt.In the initial phase of the test, the overheated, highly agitated metal melt causes intense interaction with the concrete, which leads to fast decrease of the initial melt overheat and reduction of the initially high concrete erosion rate. Thereafter, the erosion by the metal melt slows down to about 0.07 mm/s into the axial direction. Lateral erosion is by a factor of 3 smaller. Surface flooding of the melt is initiated at 800 s. Flooding does not lead to strong melt/water interactions and to penetration of water into the melt. Concrete erosion continues with about 0.040 mm/s until the melt reaches the maximum erosion limit of the crucible. Post-test analysis of the solidified melt was performed after the crucible was sectioned. The solidified melt shows no indication of water ingression from the upper surface. Tight surface crusts explain poor heat removal to the flooding water and the ongoing concrete erosion also after the top flooding.Details of the experiment are reported. The experiment shall be used for validation of models and computer codes for safety assessment.     

Experimental and theoretical investigations on the COMET concept for ex-vessel core melt retention

In the very unlikely case of a core melt accident in a nuclear power plant, the reactor pressure vessel could fail and corium melt could be released into the reactor cavity. Subsequent processes could result in a threat of the containment integrity. As a counter-measure the implementation of a core-catcher device in nuclear power plants is envisaged. Such a core-catcher concept has been developed at the Forschungszentrum Karlsruhe (FZK, Germany) within the COMET project. It is based on water injection into the melt layer from the bottom, yielding rapid fragmentation of the corium, porosity formation and thus coolability. Detailed large scale experiments with sustained heating of melts have highlighted the sequences of flooding and cooling and have been used to optimise the COMET concept. The open porosities and large surfaces that are generated during melt solidification form a porous permeable structure that is permanently filled with the evaporating coolant water and thus allows efficient short-term and long-term removal of the decay heat. Two variants of the bottom flooding concept have been developed and seem technically mature for reactor application. Corium layers up to 0.5 m high are safely arrested and cooled by water supply with 0.2 bar overpressure.The conceptual and experimental work at FZK is accompanied by theoretical investigations at IKE, University of Stuttgart. These investigations address porosity formation as well as quenching and long-term coolability of layers with resulting porosities. The aim of the theoretical work is to get a better understanding of the underlying processes of porosity formation in order to generally support the applicability of the concept for real conditions and to allow checks and optimisation for various conditions. A model for porosity formation is presented, which assumes that this process is essentially determined by strong local pressure buildup from strong evaporation due to water injection from below and the restriction of steam removal by friction in the melt. The effect of key parameters is investigated and compared to experimental results. Agreement about the influence and importance of these parameters as well as essential quantitative effects is found.     

严重事故IVR下反应堆压力容器耦合传热数值模拟分析

堆芯熔化严重事故下保证反应堆压力容器（RPV）完整性非常重要,高温蠕变失效是堆芯熔化严重事故下反应堆压力容器的主要失效模式。在进行严重事故堆芯熔化物堆内包容（IVR）下RPV结构完整性分析中,RPV内外壁和沿高度方向的温度分布以及剩余壁厚是结构分析的重要输入。本文采用CFD分析方法对RPV堆内熔融物、RPV壁以及外部气液两相流动换热进行热-固-流耦合分析,获得耦合情况下的温度场、流场、各相份额分布以及RPV的剩余壁厚,为RPV在严重事故IVR下的结构完整性分析提供依据。     

Analytical evaluation of two-phase natural circulation flow characteristics under external reactor vessel cooling

This work proposes an analytical method of evaluating the effects of design and operating parameters on the low-pressure two-phase natural circulation flow through the annular shaped gap at the reactor vessel exterior surface heated by corium (molten core) relocated to the reactor vessel lower plenum after loss of coolant accidents. A natural circulation flow velocity equation derived from steady-state mass, momentum, and energy conservation equations for homogeneous two-phase flow is numerically solved for the core melting conditions of the APR1400 reactor. The solution is compared with existing experiments which measured natural circulation flow through the annular gap slice model. Two kinds of parameters are considered for this analytical method. One is the thermal–hydraulic conditions such as thermal power of corium, pressure and inlet subcooling. The others are those for the thermal insulation system design for the purpose of providing natural circulation flow path outside the reactor vessel: inlet flow area, annular gap clearance and system resistance. A computer program NCIRC is developed for the numerical solution of the implicit flow velocity equation.     

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.     

An evaluation of a direct cooling method for the ex-vessel corium stabilization

An evaluation of the ex-vessel core catcher system of a sample advanced light water reactor was presented. The core catcher was designed to cool down the molten corium through a combined injection of water and gas from the bottom of the molten corium, which could be effective in the reduction of rapid steam generation. By using the MELCOR code, a scenario analysis was performed for a representative severe accident scenario of the ALWR, that is, the 6-in. large break loss of coolant accident without safe injection. The spreading characteristics of ejected corium at vessel breach were asymptotically evaluated on the core catcher horizontal surface. The composition of the molten corium, the decay power level, and the sacrificial concrete ablation depth with time were obtained by a sacrificial concrete ablation analysis. The corium cooling history in the core catcher during the coolant injection was evaluated to calculate the temporal steam generation rate by considering an energy conservation equation. These were used as the major inputs for the temporal calculations of containment pressure which was performed by using the GASFLOW code. Several cases with change of water and gas injection rates were calculated. It was confirmed that the bottom water/gas injection system was an effective corium cooling method in the ex-vessel core catcher to suppress the quick release of steam.     

船用堆严重事故下熔融物堆内滞留能力分析

堆芯熔化的严重事故,可能导致船用堆下封头失效、熔融物进入堆坑,危害人员及船体安全。本文采用严重事故一体化程序MAAP4,以船用堆全船断电事故为研究对象,针对低压安全注射系统投入时机、低压安全注射水流量,研究下腔室熔池形成后,投入低压安全注射系统对熔融物堆内滞留的作用。结果表明:在下腔室熔池形成后1576?s时,投入两台安全注射泵仍能有效阻止压力容器失效,实现熔融物堆内滞留;在下腔室熔池形成2646?s后,投入低压安全注射系统不能阻止压力容器失效。      

Estimation of ex-vessel steam explosion pressure loads

An ex-vessel steam explosion may occur when, during a severe reactor accident, the reactor vessel fails and the molten core pours into the water in the reactor cavity. A steam explosion is a fuel coolant interaction process where the heat transfer from the melt to water is so intense and rapid that the timescale for heat transfer is shorter than the timescale for pressure relief. This can lead to the formation of shock waves and production of missiles that may endanger surrounding structures. A strong enough steam explosion in a nuclear power plant could jeopardize the containment integrity and so lead to a direct release of radioactive material to the environment.In this article, different scenarios of ex-vessel steam explosions in a typical pressurized water reactor cavity are analyzed with the code MC3D, which is being developed for the simulation of fuel–coolant interactions. A parametric study was performed varying the location of the melt release (central, right and left side melt pour), the cavity water subcooling, the primary system overpressure at vessel failure and the triggering time for explosion calculations. The main purpose of the study was to establish the influence of the varied parameters on the fuel–coolant interaction behaviour, to determine the most challenging cases and to estimate the expected pressure loadings on the cavity walls. For the most explosive central, right side and left side melt pour scenarios a detailed analysis of the explosion simulation results was performed. The study shows that for some ex-vessel steam explosion scenarios higher pressure loads are predicted than obtained in the OECD programme SERENA phase 1.     

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.     

BETA experiments in verification of the Wechsl code: experimental results on the melt-concrete interaction

The BETA experiments are conducted to investigate the melt-concrete interaction in a large-scale melt facility using internally heated simulated core melts. The experimental findings are extrapolated to reactor accident conditions by means of computer codes verified experimentally.The experiments cover a wide range of temperatures and power rates typical of accident conditions. In high-temperature melts, fast downward erosion determines the cavity shape and the very high downward heat transfer causes the temperature of the melt to drop rapidly, even with high internal heating. Crust formation at the interface between the melt and the concrete during the low-temperature interaction allows the gases evolved by the concrete to percolate through the melt, thus establishing an effective gas driven mode of heat transfer. Measurements of gases and aerosols are reported and discussed for silicate and limestone types of 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.     

In-vessel retention of corium at the Loviisa plant

In-vessel retention of corium has been approved to be part of the severe accident management strategy for IVO's Loviisa plant. The approach selected takes advantage of the unique features of the plant such as a low power density, a reactor pressure vessel (RPV) without penetrations at the bottom, and ice-condenser containment which ensures a flooded cavity in all risk significant sequences. The thermal analyses, which are supported by an experimental program, demonstrate that, in Loviisa, the molten corium on the lower head of the RPV is externally coolable with wide margins. This paper summarizes the approach, the thermal analyses and the plant modifications being implemented.