The effect of corium composition and interaction vessel geometry on the prototypic steam explosion

This paper discusses the results of steam explosion experiments using molten material consisting of UO₂ and ZrO₂ mixture, which is called corium, to simulate a prototypic steam explosion in a nuclear reactor during a postulated severe accident. About 5–10 kg of molten material with enough superheat was poured into a pool of water in a test section at room temperature to simulate ex-vessel steam explosion in the reactor situation. Most of the experiments were externally triggered. The purpose of the experiments was to investigate the effect of material composition and average void fraction on the strength of a prototypic steam explosion, which were highlighted as major unresolved issues.The experiments were performed using two kinds of mixtures, one, corium A, at 70:30 weight percent composition of UO₂ and ZrO₂, close to eutectic composition, and the other, corium B, at 80:20 weight percent. Also, two kinds of cylindrical test sections having a different diameter were used. It turned out that corium A was likely to produce an energetic steam explosion, while corium B seldom led to an energetic steam explosion. The existence of mush phase for the non-eutectic mixture is suggested to be the reason for the difference. Comparative cross sectional views of the corium particles by scanning electron microscope supported the proposed argument. The tests performed with a narrow test section seldom led to an energetic steam explosion for both materials. An increase in average void is suggested to be the reason for the non-explosive behavior, which is consistent with the physical models employed in the current steam explosion computer codes.     

Modelling of PWR lower head failure under severe accident loading using improved shells of revolution theory

In the study of severe pressurized water reactor accidents, the scenarios that describe the relocation of significant quantities of liquid corium at the bottom of the lower head are usually investigated from the mechanical point of view. In these scenarios, the risk of a breach and the possibility of a large quantity of corium being released from the lower head exists. This may lead to an out of vessel steam explosion or to direct heating of the containment; both which have the potential to lead to early containment failure.Within the framework of the OECD Lower Head Failure (OLHF) programme, a simplified model based on the theory of shells of revolution under symmetrical loading was developed by IRSN. After successfully interpreting some other representative experiments on lower head failures, the model was recently integrated into the European integral severe accident computer ASTEC code. The model was also used to obtain the thermo-mechanical behaviour of a 900-MWe pressurized water reactor lower head, subjected to transient heat fluxes under severe accident conditions.The main objective of this paper is to present: (1) the full mathematical formulations used in the development of the model, including their matrices and integrals defined by analytical expressions; (2) the two creep laws implemented, one for the American steel SA533B1 and one for the French steel 16MND5; and (3) the various numerical interpretations of experiments using the simplified model. This paper can be considered as a theoretical manual to aid users of the simplified model during modelling of lower head failures under severe accident conditions. One of the applications presented in this paper concerns the determination of a diagram representing the vessel time to failure as a function of the pressure level and the heat flux intensity. This information has been used by IRSN in probabilistic safety assessment and severe accident management analyses.     

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。     

Semi-analytical modeling of a PWR lower head failure under severe accident conditions using an axisymetrical shell theory

The USNRC/SNL OLHF program was carried out within the framework of an OECD project. This program consisted of four one-fifth scale experiments of a reactor pressure vessel (RPV) lower head failure (LHF) under well controlled internal pressure and large throughwall temperature differentials; the objectives were to characterize the mode, timing and size of a possible PWR lower head failure in the event of a core meltdown accident. These experiments should also lead to a better understanding of the mechanical behavior of the reactor vessel lower head, which is of importance both in severe accident assessment and the definition of accident mitigation strategies. A well-characterized failure of the lower head is of prime importance for the evaluation of the quantity of core material that can escape into the containment, since this defines the initial conditions for all ex-vessel events. A large quantity of escaping corium may lead to direct heating of the containment or ex-vessel steam explosion. These are important issues due to their potential to cause early containment failure. The experiments also provide data for model development and validation. For our part, as one of the program partners, a 2D semi-analytical model has been developed and used to simulate these experiments. The aim of this effort is to develop a simplified but well predicting code that can be then implemented in European integral severe accident computer codes (ASTEC, ICARE/CATHARE). This paper presents the detailed mathematical formulation of this simplified method which is used to interpret the experimental results. The axi-symmetric shell theory under internal pressure proposed by Timoshenko has been utilised. The solution to the equilibrium equations is presented, with particular attention to the Rabotnov analytical formula. The radius and the polar angle of the deformed structure have been written as analytical expressions in order to take the large displacements and large strains into account using our mathematical formulation. The Norton type creep law and the Kachanov damage law have been used. Several failure criteria were used in the calculations and their effect on the numerical results is discussed. This 2D semi-analytical model gives very satisfactory results when compared, with the experimental and numerical results that were presented recently in the Benchmark calculations based on the first test of the OLHF program. The performance of this model is also illustrated by its capacity to accurately simulate the deformation of the lower head, including the variation of wall thickness.     

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.     

Fuel coolant interaction experiments in TROI using a UO2/ZrO2 mixture

This paper discusses the results of steam explosion experiments using reactor material carried out under “Test for Real cOrium Interaction with water (TROI)” program. About 4–9 kg of corium melt jet is delivered into a sub-cooled water pool at atmospheric pressure. Spontaneous steam explosions are observed in four tests among six tests. The dynamic pressure, dynamic load, and morphology of debris clearly indicate the cases with steam explosion. The initial conditions and results of the experiments are discussed.     

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.     

Results of recent KROTOS FCI tests: alumina versus corium melts

Recent results from KROTOS fuel-coolant interaction experiments are discussed. Five tests with alumina were performed under highly subcooled conditions, all of these tests resulted in spontaneous steam explosions. Additionally, four tests were performed at low subcooling to confirm, on one hand, the suppression of spontaneous steam explosions under such conditions and, on the other hand, that such a system is still triggerable using an external initiator. The other test parameters in these alumina tests included the melt superheat and the initial pressure. All the tests in the investigated superheat range (150–750 K) produced a steam explosion and no evidence of the explosion suppression by the elevated initial pressure (in the limited range of 0.1–0.375 MPa) was observed in the alumina tests. The corium test series include a test with 3 kg of melt under both subcooled and near saturated conditions at ambient pressure. Two additional tests were performed with subcooled water; one test was performed at an elevated pressure of 0.2 MPa with 2.4 kg of melt and another test with 5.1 kg of melt at ambient pressure. None of these tests with corium produced a propagating energetic steam explosion. However, propagating low energy (about twice the energy of the trigger pulse) events were observed. All corium tests produced significantly higher water level swells during the mixing phase than the corresponding alumina tests. Present experimental evidence suggests that the water depletion in the mixing zone suppresses energetic steam explosions with corium melts at ambient pressure and in the present pour geometry. Processes that could produce such a difference in void generation are discussed.     

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.     

Load carrying capacity of a reactor vessel head under a corium slug impact from a postulated in-vessel steam explosion

The load carrying capacity of the pressure vessel head to withstand an in-vessel steam explosion is investigated. Firstly, as a key problem, the impact of molten core material against the vessel head is studied by model experiments scaled down 1:10. Structural details are considered carefully. The results are converted to reactor dimensions using similarity theory. This approach was checked by simplified liquid-structure impact experiments in different scale. Secondly, the upward acceleration of molten core material is studied by computational models. As results the mechanical energies which the vessel head can withstand are presented.     

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%.     

Consequences of material effects on in-vessel retention

In-vessel retention (IVR) consists in cooling the corium contained in the reactor vessel by natural convection and reactor cavity flooding. This strategy of severe accident management enables the corium to be kept inside the second confinement barrier: the reactor vessel. The general approach which is used to study IVR problems is a “bounding” approach which consists in assuming a specified corium stratification in the vessel and then demonstrating that the vessel can cope with the resulting thermal and mechanical loads. Thermal loading on the vessel is controlled by the convective heat transfer inside the molten corium in the lower head. If there is no water in the vessel and if the corium pool is overlaid by a liquid steel layer, then the heat flux might focus on the vessel in front of the steel layer (“focusing effect”) and exceed the dry-out heat flux (CHF or DHF). One of the critical points of these studies is linked to the determination of the height of the molten steel layer that can stratify above the oxidic pool. The MASCA experiments have highlighted that part of molten steel may stratify under the oxidic corium which reduces the thickness of the steel layer on top of the pool. This behavior can be explained by chemical interaction between the oxide and metallic phases of the pool which confirms that these materials cannot be treated as inert species. Following these conclusions, a methodology which couples physicochemical effects and thermalhydraulics has been developed to address the IVR issue. The main purpose of this paper is to present this methodology and its application for given corium mass inventories. Attention focuses on the influence of parameters such as the ratio U/Zr and oxidation ratio of zirconia. For a 1000 MW PWR, approximately 10 t of steel stratify at the bottom of the vessel for 40% Zr oxidation, and 25 t for 30% Zr oxidation. This leads to a 25–50% increase of the mass of molten steel that is required for avoiding vessel melt-through.     

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.     

ASTEC程序中反应堆熔池结构对压力容器下封头换热计算的影响

反应堆严重事故工况下堆内环境复杂,针对下腔室内熔融物行为的试验非常有限,因此通常采用假设的熔池结构模型进行事故评价。本文使用ASTEC程序中的3种熔池结构模型,评价典型严重事故工况下不同熔池结构对下封头内壁换热及压力容器完整性的影响。计算结果表明:在外壁绝热且下封头失效仅使用温度限值的条件下,两层熔池结构导致下封头失效时间最短,且由于顶部金属层集热效应,失效位置位于熔池上部;三层熔池结构由于底层金属层的出现,使下封头下部温度持续升高而发生失效,但其失效时间长于两层熔池结构的情况。     

Fragmentation of a molten corium jet falling into water

During a severe accident of a pressurized water nuclear reactor, a large mass of corium could pour into the vessel bottom as a compact jet. When the corium mass reaches the water at the bottom of the vessel, an intense fragmentation may occur. This could lead to a significant mixing of corium and water, likely to cause a steam explosion which could damage the structures. An analytical study has been established in order to quantify the corium jet fragmentation. This study consists mainly in modeling the vapor flow surrounding the jet as well as the instability which occurs at its interface. In comparison with previous studies, this model pays particular attention to the jet-produced particles which interact with the vapor flow. A complete model has been set up in order to calculate the jet breakup length and the generated particles’ diameter under each specific situation characterized by initial conditions. This model mainly relies upon results from boundary layer theory and linear instability calculations. The full model’s results are compared to existing experiences in this field and a final correlation of the results is established. A good agreement is obtained on the jet breakup length, however the predicted particle diameter tends to be too large. This last result could be explained by a secondary breakup of the particles in water and by a large uncertainty in the vapor flow.     

Load carrying capacity of a reactor vessel head under molten core slug impact: Final report including recent experimental findings

The load carrying capacity of the pressure vessel head to withstand a postulated in-vessel steam explosion has been investigated. Former results have been confirmed and extended. For a quite conservative reference case the mechanical energy release which the head can carry turns out to be at least 1 GJ. The mechanical energy release from in-vessel steam explosions postulated can be expected to be significantly lower, and therefore will not rupture the vessel head and not endanger the containment integrity.     

Simulation of the physical effects of the interaction of molten fuel with the coolant

Conclusions The method proposed makes it possible to obtain computational estimates of the intensity of a steam explosion inside a reactor vessel and in the space below the reactor inside the melt trap. The computational investigations of the intensity of a steam explosion inside a VVéR vessel in the most likely scenario of a serious accident with efflux of melt into the bottom pressurized chamber show that under certain conditions a high pressure capable of destroying separate structural elements can develop. The mass of the interacting melt, the initial temperature, the fragmentation time, and the final size of the fragments, as well as the type of contact realized, have the greatest effect on the intensity of the steam explosion. Local steam explosions in pipes of the melt trap have a relatively low intensity and cannot have a large effect on the construction in the space below the reactor and on the containment envelope. Deceased. State Science Center of the Russian Federation — Physics and Power Engineering Institute. Translated from Atomnaya énergiya, Vol. 80, No. 1, pp. 3–10, January, 1996.     

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.     

Numerical analysis of pressure load in a PWR cavity in an ex-vessel steam explosion

Ex-vessel steam explosion may happen as a result of melting core falling into the reactor cavity after failure of the reactor vessel and interaction with the coolant in the cavity pool. It can cause the formation of shock waves and production of missiles that may endanger surrounding structures. Ex-vessel steam explosion ener- getics is affected strongly by three dimensional （3D） structure geometry and initial conditions. Ex-vessel steam explosions in a typical pressurized water reactor cavity are analyzed with the code MC3D, which is developed for simulating fuel-coolant interactions. The reactor cavity with a venting tunnel is modeled based on 3D cylin- drical coordinate. A study was performed with parameters of the location of molten drop release, break size, melting temperature, cavity water subcooling, triggering time and explosion position, so as to establish parame- ters＇ influence on the fuel-coolant interaction behavior, to determine the most challenging cases and to estimate the expected pressure loadings on the cavity walls. The most dangerous case shows the pressure loading is above the capacity of a typical reactor cavity wall.     

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.