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.     

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.     

Concept of a core cooling system and experiments performed

One of the problems which must be solved in severe accidents is the melt-concrete interaction which does occur when the core debris penetrates the lower pressure vessel head and contacts the basement. To prevent these accident consequences, a core catcher concept is proposed to be integrated into a new pressurized-water reactor design. The core catcher achieves coolability by spreading and fragmentation of the ex-vessel core melt based on the process of water inlet from the bottom.In order to justify the dominant process during flooding of the melt from the bottom, prototypic experiments with thermite melts in laboratory scale have been carried out. In these experiments flooding and early coolability of the melt is demonstrated. To obtain more detailed information on the important process of water penetration into the melt, a simulant experiment has been conducted using a transparent plastic melt with the typical viscosity behaviour of an oxidic corium melt and a temperature allowing evaporation of water. In every experiment the melt is flooded, and complete freezing in the form of a porous layer occurs within a few minutes only.     

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.     

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%.     

In-vessel corium catcher of a nuclear reactor

A new concept of an in-vessel corium melt catcher is proposed. The lower part of an elongated reactor vessel, which is filled with a sacrificial material of a proper composition, porosity, and arrangement, is used as such a catcher. The concept accounts of the scientific and design experience with the development of the ex-vessel corium catcher for the Tyan’van NPP with VVER-1000 reactors.     

Corium crust strength measurements

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

Corium spreading and coolability: CSC Project

In the context of severe accidents, large R&D efforts throughout the world are currently directed towards ex-vessel corium behaviour. Among the mitigation means which can be envisaged, the European industries and utilities are considering the implementation of a core-catcher outside the reactor pressure vessel in order to prevent basemat erosion and to stabilize and control the corium within the containment. The CSC project focused on two key phenomena for external core-catcher efficiency, reliability and safety: spreading and coolability. An experimental programme, covering different scenarios and including both simulant and real materials provided a lot of results which now constitute a large database and which enabled the qualification of computer codes.     

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.     

Results of the QUENCH-L2, DISCO-L2, and COMET-L2 experiments performed within the LACOMERA project at the Forschungszentrum Karlsruhe

The LACOMERA project at the Forschungszentrum Karlsruhe, Germany (FZK) is a 4-year action within the 5th Framework Programme of the EU which started in September 2002. Overall objective of the project is to offer research institutions from the EU Member Countries and Associated States access to four large-scale experimental facilities QUENCH, LIVE, DISCO, and COMET. These facilities are being used to investigate core melt scenarios from the beginning of core degradation to melt formation and relocation in the vessel, possible melt dispersion to the reactor cavity, and finally corium concrete interaction and corium coolability in the reactor cavity. The paper summarizes the main results obtained in the following three experiments:QUENCH-L2: boil-off of a flooded bundle. The test is of a generic interest for all reactor types, provided a link between the severe accident and design basis areas, and would deliver oxidation and thermal hydraulic data at high temperatures.DISCO-L2: fluid-dynamic, thermal, and chemical processes during melt ejection out of a breach in the lower head of a pressure vessel of the VVER-1000/320 type of reactor.COMET-L2: investigation of long-term melt-concrete interaction of metallic corium in a cylindrical siliceous concrete cavity under dry conditions with decay heat simulation of intermediate power during the first test phase, and subsequently at reduced power during the second test phase.     

压水堆严重事故下封头热斑计算

压水堆堆芯熔化事故情况下,下封头热斑会造成压力容器局部过热,导致临界热流密度发生。利用FLUENT软件对堆芯熔化事故时的下封头热斑进行计算,从流动和换热角度预测热斑导致的下封头薄弱环节。计算结果表明:堆芯熔化事故时,压力容器下封头存在两处最薄弱的位置,分别为下封头正下方正对外部冷却水位置和氧化壳与压力容器交界处。特别是在氧化壳与压力容器交界处,由于多种原因导致临界热流密度发生,使得该处熔化严重。通过设置延伸小管和附加冷却水可延迟压力容器壁面熔穿的时间。     

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.     

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.     

熔融物液柱碎化模型对先进压水堆堆外蒸汽爆炸计算的影响分析

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

增强IVR有效性的堆内注水策略研究

熔融物堆内滞留（IVR）是一项核电厂重要的严重事故管理措施,通过将熔融物滞留在压力容器内,以保证压力容器完整性,并防止某些可能危及安全壳完整性的堆外现象。对于高功率和熔池中金属量相对不足的反应堆,若下封头形成3层熔池结构,则其顶部薄金属层导致的聚焦效应可能对压力容器完整性带来更大的威胁。本文考虑通过破口倒灌及其他工程措施实现严重事故下熔池顶部水冷却,建立熔池传热模型,分析顶部注水的带热能力,建立事件树,分析顶部注水措施的成功概率及IVR的有效性。结果表明,通过压力容器内外同时水冷熔融物,能显著增强IVR措施的有效性。     

Experimental studies of oxidic molten corium–vessel steel interaction

The experimental results of molten corium–steel specimen interaction with molten corium on the ‘Rasplav-2’ test facility are presented. In the experiments, cooled vessel steel specimens positioned on the molten pool bottom and uncooled ones lowered into the molten pool were tested. Interaction processes were studied for different corium compositions, melt superheating and in alternative (inert and air) overlying atmosphere. Hypotheses were put forward explaining the observed phenomena and interaction mechanisms. The studies presented in the paper were aimed at the detection of different corium–steel interaction mechanisms. Therefore certain identified phenomena are more typical of the ex-vessel localization conditions than of the in-vessel corium retention. Primarily, this can be referred to the phenomena of low-temperature molten corium–vessel steel interaction in oxidizing atmosphere.     

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.     

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。      

Simulation of melt jet breakup and debris bed formation in water pools with IKEJET/IKEMIX

The objective of the development of the code system KESS is simulating the processes of core melting, relocation of core material to the lower head of the reactor pressure vessel (RPV) and its further heatup, modelling of fission product release and coolability of the core material. In the scope of the code development, IKEJET and IKEMIX were designed as key models for the breakup of a molten jet falling into a water pool, cooling of fragments and the formation of particulate debris beds. Calculations were performed with these codes, simulating FARO corium quenching experiments at saturated (L-28) and subcooled (L-31) conditions, as well as PREMIX experiments, e.g. PM-16. With the assumption of a reduced interfacial friction between water and steam as compared to usually applied laws, the melt breakup, energy release from the melt and pressurisation of the vessel observed in the experiments are reproduced with a reasonable accuracy. The model is further applied to reactor conditions, calculating the relocation of a mass of corium of 30 t into the lower plenum, its fragmentation and the formation of a particle bed.     

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.