Core Debris Research Articles

Investigation on multilayer failure mechanism of RPV with a high temperature gradient from core meltdown scenario

The Fukushima accident shows that in-vessel retention (IVR) of molten core debris has not been appropriately assessed, and a certain pressure (up to 8.0MPa) still exists inside the reactor pressure vessel (RPV). In the traditional concept of IVR, the pressure is supposed to successfully be released, and the temperature distributed among the wall thickness is assumed to be uniform. However, this concept is seriously challenged by reality of Fukushima accident with regard to the existence of both internal pressure and high temperature gradient. Therefore, in order to make the IVR mitigation strategy succeed, the numerical investigation of the lower head behavior and its failure has been performed for several internal pressures under high temperature gradient. According to some requirements in severe accident (SA) management of RPV, it should be ensured that the IVR mitigation takes effect in preventing the failure of the structure within a period of 72h. Subsequently, the failure time and location have to be predicted under the critical heat flux (CHF) loading condition for lower head, since the CHF is limit thermal boundary before the melt-through of RPV. In illustrating the so called ‘multilayer failure mechanism’, the structural behaviors of RPV are analyzed in terms of the stress, creep strain, deformation, damage on selected paths.

Phenomenological Analysis of Melt Progression in LWRS: Proposal of an In-Vessel Corium Progression Scenario in the Unit 1 of Fukushima Dai-ichi Nuclear Power Plant

In the field of severe accident, the description of corium progression events is mainly carried out using integral calculation codes. However, these tools are usually based on bounding assumptions because of the high complexity of phenomena. The limitations associated with bounding situations [1] (e.g., steady-state situations and instantaneous whole core relocation in the lower head) led CEA to develop an alternative approach to improve the phenomenological description of the melt progression. The methodology used to describe the corium progression was designed to cover the accidental situations from the core meltdown to the molten core–concrete interaction (MCCI). This phenomenological approach is based on the available data (including learnings from TMI-2) on physical models and knowledge about the corium behavior. It provides emerging trends and best-estimate intermediate situations. As different phenomena are unknown, but strongly coupled, uncertainties at large scale for the reactor application must be taken into account. Furthermore, the analysis is complicated by the fact that these configurations are most probably three-dimensional (3D), all the more so because 3D effects are expected to have significant consequences for the corium progression and the resulting vessel failure. Such an analysis of the in-vessel melt progression was carried out for the Unit 1 of the Fukushima Dai-ichi Nuclear Power Plant. The core uncovering kinetics governs the core degradation and impacts the appearance of the first molten corium inside the core. The initial conditions used to carry out this analysis are based on the available results derived from codes such as the MELCOR calculation code [2]. The core degradation could then follow different ways: (1) Axial progression of the debris and the molten fuel through the lower support plate, or (2) lateral progression of the molten fuel through the shroud. On the basis of the Bali program results [3] and the TMI-2 accident observations [4], this work is focused on the consequences of a lateral melt progression (not excluding an axial progression through the support plate). Analysis of the events and the associated time sequence will be detailed. Besides, this analysis identifies some number of issues. Random calculations and statistical analysis of the results could be performed with calculation codes such as LEONAR–PROCOR codes [5]. This work was presented in the frame of the OECD/NEA/CSNI Benchmark Study of the Accident at the Fukushima Dai-ichi Nuclear Power Station (BSAF) project [6]. During the years of 2012 and 2014, the purpose of this project was both to study, by means of severe accident codes, the Fukushima accident in the three crippled units, until 6 days from the reactor shutdown, and to give information about, in particular, the location and composition of core debris.

FISSION-PRODUCT RELEASES FROM A PHWR TERMINAL DEBRIS BED

During an unmitigated severe accident in a pressurized heavy water reactor (PHWR) with horizontal fuel channels, the core may disassemble and relocate to the bottom of the calandria vessel. The resulting heterogeneous in-vessel terminal debris bed (TDB) would likely be quenched by any remaining moderator, and some of the decay heat would be conducted through the calandria vessel shell to the surrounding reactor vault or shield tank water. As the moderator boiled off, the solid debris bed would transform into a more homogeneous molten corium pool located between top and bottom crusts. Until recently, the severe accident code MAAP-CANDU assumed that unreleased volatile and semi-volatile fission products remained in the TDB until after calandria vessel failure, due to low diffusivity through the top crust and the lack of gases or steam to flush released fission products from the debris. However, national and international experimental results indicate this assumption is unlikely; instead, high- and medium-volatility fission products would be released from a molten debris pool, and their volatility and transport should be taken into account in TDB modelling. The resulting change in the distribution of fission products within the reactor and containment, and the associated decay heat, can have significant effects upon the progression of the accident and fission-product releases to the environment. This article describes a postulated PHWR severe accident progression to generate a TDB and the effects of fission-product releases from the terminal debris, using the simple release model in the MAAP-CANDU severe accident code. It also provides insights from various experimental programs related to fission-product releases from core debris, and their applicability to the MAAP-CANDU TDB model.

Pressure drop and average void fraction measurements for two-phase flow through highly permeable porous media

The modeling of pressure drop for two-phase flows through porous media is a key point to assess the coolability of debris beds resulting from nuclear severe accidents. Models involve several parameters which are non-linear functions of the void fraction, e.g. relative permeabilities. Their identification requires that experimental data include the measurement of void fraction. This paper presents a new technique developed to reach this objective. The method is based on the use of a capacitance probe and has been validated by comparison with a weighing method. The validation has shown that the accuracy is better than 10%. The measurement device has been implemented in the CALIDE facility, at IRSN, which has been designed to perform air–water flow through debris bed. Tests have been carried out with beds made of single size 4mm and 8mm beads. Measurements of pressure drop and average void fraction are reported in the paper, for air and water flow rates representative of flows that would result of either the reflooding of the damaged core or the cooling of corium debris in a stagnant pool of water. Finally, the pressure drop models used in severe accident simulation codes, based on generalizations of the single-phase Ergun law, have been assessed against the new data. It has been observed that generalized Ergun laws including an interfacial drag term accurately predict the pressure drop and the void fraction for flows with a zero net water velocity.

Hydrogen risk for advanced PWR under typical severe accidents induced by DVI line break

Advanced passive pressurized water reactor (APWR) relies on In-Vessel Retention (IVR) of molten core debris under severe accidents, which makes hydrogen generation and distribution different with other PWRs. Hydrogen generation and hydrogen risk for APWR is studied with the integrated severe accident plant model, including Reactor Coolant System (RCS), engineered safety features (ESFs), simplified secondary side, containment and passive containment cooling system (PCCS). Initial events of DVI line break with passive core cooling system (PXS) compartment not flooded (named Case 1), DVI line break with PXS compartment flooded through broken DVI line (named Case 2), and DVI line break with IRWST direct injection available based on Case 1 (named Case 3) are selected to study hydrogen generation in pressure vessel, distributions in containment, and hydrogen risk controlled with igniters. Results show that hydrogen generated in pressure vessel is 257kg, 461kg and 499kg for Case 1 to Case 3 due to different thermal hydraulic characteristics. During maintaining the water level in the pressure vessel with the safety measures, zirconium-steam reaction is violent and the duration of hydrogen generation is longer than the other two cases due to the availability of plenty of steam, resulting in an amount of hydrogen generation, and hydrogen distribution in upper containment is nearly 10%. Case 3 is screened to analyze the effectiveness of hydrogen control system consisting of igniters, showing that the hydrogen control system can mitigate hydrogen risk when igniters are available at the SAMG inlet condition.

Experimental simulation of fragmentation and stratification of core debris on the core catcher of a fast breeder reactor

The complex and coupled phenomena of two simultaneous molten metal jets fragmenting inside a quiescent liquid pool and settling on a collector plate are experimentally analysed in the context of safety analysis of a fast breeder reactor (FBR) in the post accident heat removal phase. Following a hypothetical core melt down accident in a FBR, a major portion of molten nuclear fuel and clad/structural material which are collectively termed as ‘corium’ undergoes fragmentation in the bulk coolant sodium in the lower plenum of the reactor main vessel and settles on the core catcher plate. The coolability of this decay heat generating debris bed is dependent on the particle size distribution and its layering i.e., stratification.Experiments have been conducted with two immiscible molten metals of different densities poured inside a coolant medium to understand their fragmentation behaviour and to assess the possibility of formation of a stratified debris bed. Molten aluminium and lead have been used as simulants in place of molten stainless steel and nuclear fuel to facilitate easy handling. This paper summarizes the major findings from these experiments. The fragmentation of the two molten metals are explained in the light of relevant dimensionless numbers such as Reynolds number and Weber Number. The mass median diameter of the fragmented debris is predicted from nonlinear stability analysis of slender jets for lead jet and using Rayleigh's classical theory of jet breakup for aluminium jet. The agreement of the predicted values with the experimental results is good. These experiments indicate that complete segregation of the lighter component above the denser component does not happen. However, there is segregation observed in the radial direction to a certain extent in the case of rib-partitioned collector plate pointing to the significance of the geometry of the collector plate in settling and segregation aspects. It is observed that fragmentation phenomenon with its interplaying governing forces, interferes with the settling aspects and seems to suppress stratification that arises due to density differences between the molten melts. Moreover, the thermophysical properties of the metals play a crucial role in the fragmentation and settling behaviour as exemplified by the relevant dimensionless numbers.

Creep deformation and damage behavior of reactor pressure vessel under core meltdown scenario

The Fukushima accident shows that In-Vessel Retention (IVR) of molten core debris has not been appropriately assessed, and a certain pressure (upto 8.0 MPa) still exists inside the reactor pressure vessel (RPV). Generally, the pressure was supposed to be successfully released, and the externally cooled lower head wall mainly experienced the temperature difference which may be more than 1000 °C. Therefore, in order to make the IVR succeed, it is necessary to investigate the creep behavior and damage distribution of the RPV under complex thermal-mechanical loadings. Accordingly, considering the unlikely core meltdown scenario as a severe accident condition, the failure site and time have to be predicted for a safety assessment purpose. Due to heavy thick wall of the RPV, the high temperature gradient inevitably results in various failure modes, i.e., plastic failure and creep failure. In disclosing it, the finite element model (FEM) has been developed for simulating the failure processes and the visco-plastic behavior of vessel wall. In accounting for the most dangerous situation, the critical heat flux (CHF) loading is applied on the RPV as a limit boundary. Subsequently, some failure modes has been found as local melting, plastic damage and creep damage by FEM. Furthermore, it is found that the RPV failure varies with the internal pressures. For the situation of low pressure and high temperature, the failure dominated by creep takes place at the hot focus site. With the further increase of pressure, the thinnest wall thickness is weakest for maintaining the RPV integrity, the damage of which initiates at the outside of RPV by plasticity. Besides, the multiaxial state of stress is found to accelerate the damage evolution on the thinnest site, and the uniaxial failure criterion underestimate the damage accumulation.

Development of an ex-vessel corium debris bed with two-phase natural convection in a flooded cavity

During severe accidents of light water reactors (LWRs), the coolability of relocated corium from the reactor vessel is a significant safety issue and a threat to the integrity of containment. With a flooded cavity, a porous debris bed is expected to develop on the bottom of the pool due to breakup and fragmentation of the melt jet. As part of the coolability assessment under accident conditions, the geometrical configuration of the debris bed is important. The Debris Bed Research Apparatus for Validation of the Bubble-Induced Natural Convection Effect Issue (DAVINCI) experimental apparatus facility was constructed to investigate the formation of debris beds under the influence of a two-phase flow induced by steam generation due to the decay heat of the debris bed. Using this system, five kilograms of stainless steel simulant debris were injected from the top of the water level, while air bubbles simulating the vapor flow were injected from the bottom of the particle catcher plate. The airflow rate was determined based on the quantity of settled debris, which will form a heat source due to the decay of corium. The radial distribution of the settled debris was examined using a ‘gap–tooth’ approach. Based on the experimental results of this study, an analytical model was established to describe the spreading of the debris bed in terms of two-phase flow and the debris injection parameters. This model was then used to analyze the formation of debris beds at the reactor scale. A sensitivity analysis was carried out based on key accident parameters, including the quantity of corium melt, cavity flooding level, volumetric decay heat rate, and the size of the melt jet.

球状多孔質体とナノ流体を用いた飽和プール沸騰限界熱流束の向上

One strategy for severe accidents is in-vessel retention (IVR) of corium debris. IVR consists of external cooling of the reactor vessel in order to remove decay heat from the molten core through the lower head of the vessel. However, heat removal is limited by the occurrence of the critical heat flux condition (CHF) at the outer surface of the reactor vessel. Therefore we proposed the CHF enhancement technique in a saturated pool boiling by attachment of a honeycomb porous plate (HPP) on the heated surface. However, the reactor vessel to install HPP has curvature. The CHF enhancement depend on how HPP are attached firmly to the heated surface, so it is important to establish the method to place the HPP on the surface of reactor vessel with curvature. Therefore, we propose the way using porous cellulose beads and nanofluid. As a result, the CHF is shown to be enhanced double compared with that of a plain surface of pure water.

多核種高除染性空気浄化システム開発による作業被曝低減化研究

多核種高除染性空気浄化システム開発による作業被曝低減化研究

MELCOR 2.1 analysis of melt behavior in a BWR lower head during LOCA and SBO accident

Melt behavior in reactor pressure vessel (RPV) lower head is important for estimating melt leakage to reactor cavity (pedestal), molten core concrete interaction (MCCI) and radioactivity release to the environment. In order to expand universal knowledge of severe accident research on boiling water reactors (BWRs), the melt behavior in BWR lower head and leakage paths during hypothetical loss-of-coolant accident (LOCA) and station blackout (SBO) accident were analyzed with MELCOR 2.1. The instrument guide tube (IGT) failure, a potential lower head failure mechanism, was modeled in MELCOR code. When IGTs failure was not considered the lower head failed through creep rupture, resulting in the discharge of all the debris into the reactor cavity. Otherwise, the debris discharged out through the openings at the IGT locations, but a part of debris still remained in the lower plenum as solid state. The pressure vessel wall did not suffer from creep rupture after IGTs failure. The timing of events in SBO accidents were delayed compared with LOCAs, and the oxidation degree of core debris was higher in SBO accidents. In each case, only SS (stainless steel) and a small amount of Zr (zircaloy) were present in molten state in the lower plenum. Most of the discharged debris was in solid state.

Experimental investigation of particulate debris spreading in a pool

Termination of severe accident progression by core debris cooling in a deep pool of water under reactor vessel is considered in several designs of light water reactors. However, success of this accident mitigation strategy is contingent upon the effectiveness of heat removal by natural circulation from the debris bed. It is assumed that a porous bed will be formed in the pool in the process of core melt fragmentation and quenching. Debris bed coolability depends on its properties and system conditions. The properties of the bed, including its geometry are the outcomes of the debris bed formation process. Spreading of the debris particles in the pool by two-phase turbulent flows induced by the heat generated in the bed can affect the shape of the bed and thus influence its coolability.The goal of this work is to provide experimental data on spreading of solid particles in the pool by large-scale two-phase flow. The aim is to provide data necessary for understanding of separate effects and for development and validation of models and codes. Validated codes can be then used for prediction of debris bed formation under prototypic severe accident conditions. In PDS-P (Particulate Debris Spreading in the Pool) experiments, air injection at the bottom of the test section is employed as a means to create large-scale flow in the pool in isothermal conditions. The test section is a rectangular tank with a 2D slice geometry, it has fixed width (72mm), adjustable length (up to 1.5m) and allows water filling to the depth of up to 1m. Variable pool length and depth allows studying two-phase circulating flows of different characteristic sizes and patterns. The average void fraction in the pool is determined by video recording and subsequent image processing. Particles are supplied from the top of the facility above the water surface. Results of several series of PDS-P experiments are reported in this paper. The influence of the gas flow rate, pool dimensions, particle density and size on spreading of the particles is addressed. A preliminary scaling approach is proposed and shown to provide good agreement with the experimental findings.

Empirical closures for particulate debris bed spreading induced by gas–liquid flow

Efficient removal of decay heat from the nuclear reactor core debris is paramount for termination of severe accident progression. One of the strategies is based on melt fragmentation, quenching and cooling in a deep pool of water under the reactor vessel. Geometrical configuration of the debris bed is among the important factors which determine possibility of removing the decay heat from the debris bed by natural circulation of the coolant. For instance, a tall mound-shape debris bed can be non-coolable, while the same debris can be coolable if spread uniformly. Decay heat generates a significant amount of thermal energy which goes to production of steam inside the debris bed. Two-phase flow escaping through the top layer of the bed becomes a source of mechanical energy which can move the particulate debris along the slope of the bed. The motion of the debris will lead to flattening of the bed. Such process is often called “self-leveling” phenomenon. Spreading of the debris bed by the self-leveling process can take significant time, depending on the initial debris bed configuration and other parameters. There is a competition between the time scales for reaching (i) a coolable configuration of the bed, and (ii) onset of dryout and re-melting of the debris. In the previous work we have demonstrated that the rate of particulate debris spreading is determined by local gas velocity and local slope angle of the bed. In this work we develop a scaling approach and a closure for prediction of debris spreading rate based on generalization of available experimental data. We demonstrate that introduced scaling criteria are universal for particles of different shapes and size distributions.

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.

The influences of Pu and Zr on the melting temperatures of the UO2–PuO2–ZrO2 pseudo-ternary system

Specimens of (U, Pu, Zr)O2 were prepared as simulated corium debris that were assumed like debris generated in the severe accident of the Fukushima Daiichi Nuclear Power Plant and their melting temperatures were measured by the thermal arrest technique in order to evaluate the influence of plutonium and zirconium content on the melting temperature of the corium debris. From the evaluation, it was found that the influence of zirconium on the melting temperatures of both (U, Pu, Zr)O2 and (U, Zr)O2 was similar and that the melting temperature of (U, Pu, Zr)O2 had a local maximum value in the Pu-content between 0 and 20 mol%. The UO2–PuO2–ZrO2 pseudo-ternary phase diagram at 2900 and 3000 K was evaluated from the present experimental results and previously reported results.

Experimental analysis of heaping and self-levelling phenomena in core debris using lead spheres

The phenomena of core debris heaping and self-levelling on the core catcher plate are important from cooling considerations and also from recriticality aspects to ensure the safety of the reactor main vessel after the accident. A series of experiments is conducted in water, with funnel guided pouring of lead spheres on to a flat collector plate to mimic the heaping behaviour of core debris in sodium on core catcher plate. The rationale behind the choice of Lead–water pair is that their density ratio is equal to that between the nuclear fuel and the coolant in the reactor. From dimensional analysis it is shown that that repose angle of the heap (θ) is a function of particle Reynold’s number (Re), Froude number (Fr) and Archimedes number (Ar) apart from density ratio and linear dimension ratios. Using the experimental data obtained from systematic parametric studies, the functional dependence of repose angle on Re, Fr and Ar is determined and the following empirical correlation is developed. θ=26Re-0.38Fr0.15Ar0.20 In the nuclear reactor, the decay heat generated within the core debris leads to natural convection and then local boiling within the coolant because of which self-levelling of the heap is expected to occur. To understand the evolution of repose angle with time, experiments have been conducted with bottom heated lead spheres, in the second part of the study. About 25% reduction of repose angle is observed in the first half an hour of heating in the natural convective regime itself which serves to reduce the possibility of coolant dryout occurring within the debris bed. But perfect levelling leading to a plane level is not achieved in the experiments with lead spheres.

Creep and Damage Analysis of Reactor Pressure Vessel Considering Core Meltdown Scenario

The Fukushima accident shows that In-Vessel Retention (IVR) of molten core debris has not been appropriately assessed, and a certain pressure (up to 8.0MPa) still exists inside the reactor pressure vessel (RPV). Generally, the pressure is supposed to successfully be released, and the externally cooled lower head wall mainly experiences the temperature difference which may be more than 1000°C. Therefore, in order to make the IVR succeed, it is necessary to investigate the creep behaviour and damage distribution of the RPV under complex thermal-mechanical loadings. Accordingly, considering the unlikely core melt down scenario for a light water reactor (LWR) a possible failure mode of the reactor pressure vessel (RPV) and its failure time has to be predicted for a determination of the loadings on the containment. Due to the thickness of RPV, the high temperature gradient results in various failure modes, i.e., plastic failure and creep failure. In disclosing the failure mechanism, the finite element model has been developed simulating the thermal processes and the visco-plastic behaviours of vessel wall. An advanced model for creep damage has been established to analyze the fracture time and fracture position of a vessel with an internally heated melt pool. Before the above, the stress and strain distributions along the wall thickness are investigated by ABAQUS software. Finally, the result shows that the calculated stress outside the RPV is lower than the yield stress of the material through most thickness. It is concluded that the RPV can maintain its integrity under IVR with given time, even if there exists the internal pressure of 8MPa.

B244 安全向上策に関わる技術課題3 : 圧力容器の健全性(OS3 軽水炉・新型炉・原子力安全(6))

Core damage and melt progression behavior due to loss of core cooling function is one of the important factors in the severe accident. After TMI-2 accident, some integral severe accident analysis codes have been developed; however, the core damage behaviors, the relocation behavior of molten core debris and failure modes of reactor pressure vessel have still large uncertainty. The Atomic Energy Society of Japan (AESJ) developed a new thermal-hydraulics safety evaluation fundamental technology enhancement strategy roadmap (TH-RM) for LWR safety improvement and development after Fukushima Daiichi NPP accident. This paper describes the major technical issues and activities to secure the integrity of reactor pressure vessel which are discussed in the TH-RM developing process.

ICONE23-1053 OPPORTUNITIES AND NEED TO FURTHER UPGRADE CANDU REACTORS FOR IMPROVED SEVERE ACCIDENT MITIGATION AND REDUCTION OF RISK

Independent deterministic analyses of system response and severe accident progression after a station blackout scenario for CANDU single and multi-unit reactors have unveiled a number of design vulnerabilities that cause uncontrolled pressure boundary ruptures; premature expulsion of coolant from main loops and from the moderator heat sink; direct exposure of core debris and fission product releases to the containment; thermo-mechanical failure of the thin shell Calandria vessel welds; accelerated production of hydrogen with containment boundary failures by steaming as well as by sparsely populated PARS units potentially exposed to high concentration deuterium/hydrogen they are unable to adequately mitigate. A number of design enhancements can however be undertaken to minimize risk from a severe accident by eliminating or avoiding some of the undesirable system responses. Engineered design measures to also avoid some the undesirable accident progression paths have been summarized in the paper. The Canadian regulatory body CNSC prepared a number of design enhancement requests as part of post Fukushima reviews but then accepted measures and submissions that do not meet the public expectations for risk reduction after Fukushima on all counts. Without thoughtful and timely design changes, consequences of a severe core damage in a CANDU reactor can pose risks that are still unacceptable especially after the hype that surrounded the utility 'Stress Tests' and regulatory 'Action Items' that followed Fukushima disaster in 2011. It is stressed that operating CANDU and other similar PHWRs in India are unique; have an enviable safety record; have maintained very high operational loads and the observations in this paper relate to requirements for mitigating a range of potential severe accidents whose probability has been demonstrated to be relatively high given that less than 15000 reactor years of operation have resulted in 3 severe accidents that have caused more off-site damage than ever anticipated. That is something the founding fathers did not consider within the design basis and the current regulating bodies have a difficult time wrapping their head around. Utilities that operate the plants have shown no visible leadership in self assessments; have rejected any attempts to enlighten them and the risk posed by continued operation of these plants can only be minimized by an open discussion of the totality of issues as attempted in this paper.

Numerical analysis of freezing controlled penetration behavior of the molten core debris in an instrument tube with MPS

The freezing controlled penetration behavior of the molten core debris in the BWR instrument tube had been analyzed with a two dimensional geometry under severe accident conditions using Moving Particle Semi-implicit (MPS) method. The change of melt viscosity with temperature in the phase transition region was taken into account in the present MPS method. Accordingly, in order to improve the computation speed of such highly viscous fluid, the implicit calculation scheme was employed to solve the viscous term in this study. The surface tension model based on the inter-particle potential force was incorporated in the MPS method to track the melt leading front more accurately. The present MPS method was validated first by simulating the experiment of molten aluminum oxide penetrating in a prototypical PWR instrument tube which was performed by EPRI. The comparison of the predicted penetration length and the measured results showed a good agreement selecting a parameter. Then the penetration and solidification behaviors of molten stainless steel and uranium dioxide in the BWR instrument tube were simulated under a wide parametric range. The computational results showed that the melt penetration length increased with the melt superheat, and the melt had plugged the tube in all simulations. The melt flow resistance increased due to the formation of the crust on the tube surface and the increase of melt viscosity in the phase transition region. The present results indicated that the typical melt penetration and solidification behavior in the instrument tube was successfully revealed by MPS method.