Vessel Lower Head Research Articles

Development of axisymmetric shell model for creep deformation analysis of reactor pressure vessel lower head

Under severe accident conditions of nuclear power plants, the lower head of a reactor pressure vessel can be at risk of creep deformation and potential failure. This study presents a computational model that can represent the creep deformation of the lower head through the application of shell theory. To account for the large deformation due to creep, kinematic equations are derived and implemented. Particularly, this study provides details of the mathematical formulation, which were lacking in previous studies. The analysis results using the developed model are compared with those of finite element analysis, considering deformation history, stress distribution, and deformed shape. A dimensionless time for the creep deformation is proposed based on the results, which can characterize the effective strain at failure. Finally, the developed model is applied to OECD lower head failure test for validation.

TARGET – Development of Submersible ROV System for BMN Inspection

Most PWRs have penetrations in the RPV lower heads for in-core nuclear instrumentation. These penetrations generally are made of nickel-based Inconel Alloy 600. Weld materials are typically Alloy 82/182. Operating conditions of PWR plants are causing nickel-based alloys cracking through a process called primary water stress corrosion cracking (PWSCC). In 2003, the licensee for the South Texas Project Unit 1 (STP-1) identified apparent boron deposits on the lower RPV head near two bottom mounted nozzles (BMNs). The NRC issued Bulletin 2003-02 to obtain information on licensee inspection activities and inspection plans for the RPV lower head. EPRI issued MRP-206 report that provides inspection and evaluation guidelines for BMNs for PWR plants, including guidelines for periodic bare metal visual examination for evidence of primary coolant leakage, or periodic non-visual nondestructive examinations for indications of service-induced cracking. The non-visual inspections (ultrasonic testing examination) may detect service-induced degradation before through-wall cracking, leakage, circumferential cracking below the bottom of the J-groove weld, release of loose parts, or incipient boric acid wastage of the low-alloy steel reactor vessel lower head material occurs. Therefore, periodic examinations will adequately manage potential for cracking by PWSCC and preserve structural integrity. INETEC developed TARGET system for BMNs inspection, consisted of submersible ROV and specially designed probe, composed of several UT probes. UT system and technique to detect, length and depth size the service-induced degradation in the BMN volume material is developed. The EPRI NDE Center performed a technical review and validated INETEC’s ultrasonic examination technique for BMNs. Aforementioned validation was done according to requirements defined by: 1) MRP-206, 2) MRP-411.

Numerical prediction of slug flow boiling heat transfer in the core-catcher cooling channel for severe accident mitigation in nuclear power plant

This paper presents the steps followed to implement and validate a hybrid multiphase flow model in the open-source code, OpenFOAM. The modeling approach couples the interface capturing model with the dispersed flow model. The resulting multiphase model can be used to predict the slug flow boiling regime. The flow regime in question occurs during the external cooling of a core-catcher and in-vessel retention (IVR) which are severe accident mitigation strategies. A distinctive key feature of this multiphase-type flow is the co-existence of large-scale slug vapor bubbles with both dispersed vapor bubbles and the carrying liquid phase. The slug vapor bubbles are generated from the coalescence of the smaller dispersed bubbles. Also, due to the tilted orientation of the core-catcher and reactor vessel lower head (for the IVR option), these large-scale bubbles remain in the vicinity of the heated surface, while being transported by the flow. This is due to the buoyancy force acting upward in these two design configurations. The latter phenomenon engenders the fact that a liquid film is occupying a thin layer separating the large bubbles from the heated surface. Under such flow conditions, the existing wall boiling model, commonly known as the (Rensselaer Polytechnic Institute) RPI model, has been demonstrated to be inadequate for the determination of the boiling heat transfer characteristics. Therefore, an extended near-wall boiling model accounting for the conduction heat flux across the liquid film (trapped underneath the slug bubbles) is formulated and implemented in this study. Using this enhanced model, the simulation of a slug flow boiling on a downward-facing heated surface produces a better prediction of the wall superheat than the original model. In addition, the morphologies of the vapor slug coexisting with dispersed bubbles are adequately captured and compared fairly well with experimental visualizations. This new multiphase model is then used to simulate a prototypical core-catcher cooling channel. Once again, a fair representation of the wall heat transfer is predicted in good agreement with measurements. Finally, it has been also successfully proven that under subcooled nucleate flow boiling conditions, the present model can reproduce the RPI model predictions.

Mass transfer between a continuous oxide phase (U-O-Zr) and a steel droplet at liquid state: Potential impact on corium pool behaviour during in-vessel melt retention

In the frame of light water reactors’ severe accidents (SA), the success of a mitigation based on the core melt retention in the vessel lower head largely depends on the corium pool phases stratification. Such a melt is composed of molten partially oxidised materials relocated from the core and molten steel mainly coming from the ablation of lower head internal structures and the vessel wall during the transient. Because of the miscibility gap in the U-O-Zr-steel thermodynamic system at liquid state, such a pool can go through various stratified states during a transient with a heavy metal phase at the bottom or a light metal one at the top.Actually, there is a significant knowledge gap related to steel migration to the pool that influences the stratification transient and can lead in some cases to a large uncertainty on the heat flux that the pool imposes on the vessel lower head. In particular, when the ablated vessel wall is considered, the molten steel relocation, either as a continuous lighter phase above the pool or as dispersed phase in the oxide pool, depends on the behaviour of the refractory crust at the pool/vessel interface (that can undergo dissolution or mechanical damage). In the case of a dispersed metal phase in the oxide pool, determining the subsequent state of the lower head pool requires the evaluation of the thermochemical interaction between both materials and their possible hydrodynamic separation. To the best of our knowledge, no model described these coupled phenomena in integral SA codes.In this paper, an integral model is proposed to better assess the phenomena at stake in such a case and clarify the need (or not) to enhance stratified pool models in SA codes. To do so, this integral model was constructed on “first-order” hypotheses; in particular, thermal equilibrium is assumed and a simplified representation of the U-O-Zr-steel thermodynamic system is considered. The model has been implemented in the PROCOR software platform and used for a parametric analysis that provides insightful order of magnitudes and trends. The main parameters are the droplet size, the corium chemical composition and the model closures associated with interfacial mass transfer.The results reported here show that mass transfer, “competing” with droplet hydrodynamics, plays an important role for droplets of millimetric size that can, depending on the pool composition, relocate under the oxide under the form of a continous heavy metal layer. For larger droplets, it is a light metal layer above the pool that can be expected. The prolonged existence of a dispersed metal phase that would remain mixed in the oxide phase seems to be only possible for a very narrow range of pool composition near the stratification inversion threshold.

Three-dimensional numerical simulation of droplet formation by Rayleigh–Taylor instability in multiphase corium

During a severe accident in a nuclear reactor, the melting of the core may lead to the formation of a multiphase liquid pool (corium) in the vessel lower head. The heat transfer at the boundary with the vessel is affected by diffusive and convective mass fluxes. In particular, the development of Rayleigh–Taylor instabilities influence the thickness of the top metallic layer and therefore the “focusing effect” of the heat flux, which is the main risk for the vessel integrity. We use a Cahn–Hilliard pseudo-binary model to describe the uranium/oxygen/zirconium/iron mixture. The diffusion and the convection are governed by the Cahn–Hilliard equation and the Navier–Stokes equations under the Boussinesq approximation. In this work, the model is isothermal and the buoyancy force is only due to the gradient of chemical composition. The model is solved in three dimensions with a pseudo-spectral code. The initial configuration consists of a light layer of iron-rich fluid above a heavy layer of uranium/oxygen/zirconium mixture. A thin layer of heavier metallic phase lays at the interface and eventually triggers a Rayleigh–Taylor instability. The metallic phase forms a plume which falls downward and then breaks up into droplets due to the Rayleigh-Plateau instability. The phenomenon is alimented by diffusion which generates the heavy metallic phase at the interface. The droplet formation observed in an experiment of corium stratification transient from the literature is qualitatively captured. The mobility, the viscosity and the surface tension are shown to have an influence on the mass transfer.

Investigation of the pressure vessel lower head potential failure under IVR-ERVC condition during a severe accident scenario in APR1400 reactors

In the event of a core meltdown in a high-power reactor, the integrity of the reactor pressure vessel is presumably protected by severe accident mitigation systems such as in-vessel retention external reactor vessel cooling (IVR-ERVC). However, in the late phase of the accident, two possible locations on the RPV are prone to failure: the location of the focusing effect and location of in-core instrument penetration. These two potential points of damage in the RPV are investigated in this study. A numerical model for the prediction of the natural convection, melting, and solidification processes for IVR-ERVC is presented. The model is based on the enthalpy-porosity approach with an extension for continuous liquid fraction function. The model is implemented in open-source field operation and manipulation (OpenFOAM) computational fluid dynamic code to produce a new solver which is based on the combination of conjugate heat transfer solver and buoyant-driven natural convection solver and the new solver is validated against the melting Gallium experimental test, in-core instrumentation failure experimental test, and BALI experimental test. This numerical model is applied for the investigation of the RPV rupture at the location of the focusing effect and in-core instrumentation penetrations. Severe ablations of the cladding and the weld materials are observed at a heat load of about ~1800 K which is expected to lead to the ejection of the penetration tubes if the force holding the penetration tube in place is lower than the force exerted by the system pressure. Subsequently, a two-layer IVR configuration is assessed and the integrity of the RPV is found not to be compromised under external reactor vessel cooling. However, in the case of a boiling crisis, the temperature of the ex-vessel wall is expected to rise quickly and this is simulated by increasing the ex-vessel wall temperature. The RPV is found to fail near the beltline due to a phenomenon known as focusing effect when the ex-vessel wall temperature rises above 1200 K.

Analysis on CHF deviation caused by geometry of the heater blocks in a validation test of IVR-ERVC technology

The innovative technology of In-Vessel Retention – External Reactor Vessel Cooling (IVR-ERVC) is a mitigation countermeasure for severe accidents, which has been applied in the design of advanced pressurized water reactors. The validity of IVR-ERVC is restricted to boiling crisis and critical heat flux (CHF) of the reactor vessel lower head. A cylindrical slice heater block was usually employed to simulate the reactor vessel lower head in the majority of existing experiments, which may cause some uncertainties in CHF under natural circulation conditions. In the present paper, an experiment was constructed to study the characteristics of natural circulation and critical heat flux. Apart from the conventional cylindrical slice, a spherical slice was also employed as the heater block. The experimental data indicated that the natural circulation flow rate increases with increasing inlet water temperature, and shows large oscillations with the lift of circulation height. Besides, CHF decreases with the increase of inlet temperature. It was found that the CHF in the spherical slice is lower than that in the cylindrical slice from 10.3% to 28.4%, depending on the inlet water temperature. For the first time, the CHF uncertainty caused by the geometrical difference of the heater block was clarified.

Numerical simulations of a creep rupture damage in Surry PWR primary circuit under TMLB accident

In the nuclear safety field, the thermomechanical phenomena that occur during an accident in a nuclear power plant (NPP) are of particular importance. During some reactor severe accidents, the decay energy of the core is transferred via the reactor cooling system (RCS) to other parts of this circuit. The associated heatup of RCS structures can lead to pressure boundary failures; with notable vulnerabilities, i.e., the lower head vessel, the pressurizer surge line, the hot leg nozzles, etc. The potential for a primary circuit component rupture is of particular concern because fission products could be released into the environment by such a failure. The risk is higher in-vessel; the relocated debris toward the lower head vessel, following a potential melting of the core, leads to the rupture of the most important protective barrier. Primary circuit structures are subject to the effects of high pressure and high temperature throughout their service life. Such loadings generally lead to creep rupture. This risk is more likely when these structures experience transients or severe accidents. One of the most studied accidents is the station blackout (SBO). SBO without operator actions accident (TMLB’ sequence) is considered as one of the most likely scenarios that may threaten the integrity of vulnerable RCS pressure boundaries. In this manuscript, we have evaluated, using the system code RELAP5/SCDAPSIM 3.4, the damage caused by rupture of some structures of the primary circuit of the Surry NPP. The structures analyzed are the surge line, the hot leg nozzles and the lower head vessel. RELAP5/SCDAPSIM 3.4 results indicate that surge line (hot leg) failures will be the first failures in the RCS pressure boundary.

Status of computational fluid dynamics for in-vessel retention: Challenges and achievements

During a severe accident in a nuclear reactor, core damage occurs and may lead to the formation of corium, followed by relocation to the vessel lower head. The decay heat released by the corium can threaten the integrity of the vessel, if no effective cooling mechanism is in place. In-Vessel Retention (IVR) is a severe accident mitigation strategy that has been shown to work for low-to-intermediate power reactors. For high power reactors, many uncertainties still exist. In an attempt to remove some of these uncertainties, the European H2020 IVMR project was launched in 2015. The focus of this project is on obtaining the additional, necessary, experimental data in order to improve on current modelling strategies. One of the modelling strategies investigated is the potential use of CFD codes in assessing the feasibility of IVR for high power reactors. The main focus of the CFD studies is on two important aspects of IVR: the presence of a metallic layer on top of the corium pool and the homogenous corium pool. These aspects are analysed by studying the thermal hydraulic features of a thin metal layer and that of a homogeneous pool. In this paper, first the used codes and numerical approaches are presented. The numerical models are subsequently assessed by comparing numerical results with relevant simulant-based experimental data, resulting, in general, in good agreement. The codes are then used to perform exploratory computations under prototypical conditions. While the behaviours of water and prototypical materials are similar for the oxide pool, significant differences are observed for the metallic layer.

Sensitivity and uncertainty analysis of the vessel lower head failure mode and melt release conditions in Nordic BWR using MELCOR code

Nordic boiling water reactors (BWR) employ ex-vessel debris coolability as a severe accident management (SAM) strategy. Melt release mode into a deep water pool located in the lower drywell is a major source of uncertainty for success of such strategy. The main goal of this work is to identify the major contributors to the uncertainty in prediction of vessel failure mode, timing and melt release conditions in Nordic BWR using MELCOR severe accident analysis code. There is a forest of control rod guide tubes (CRGTs) and instrumentation guide tubes (IGTs) in the lower head of a BWR. Failure of the penetrations is considered in this work along with the gross creep rupture of the vessel lower head. Modelling of penetrations failure in MELCOR code is based on parametric models, allowing a user to select different assumptions, such as temperature threshold for penetrations failure, modelling of penetrations assembly and respective failure modes, mode of solid and liquid debris ejection from the vessel. In this work we perform sensitivity analysis to the MELCOR modelling options and sensitivity parameters in unmitigated station blackout scenario (SBO). Results of analysis suggest that (i) vessel breach due to penetration failure is predicted to occur before creep-rupture failure of the vessel lower head, (ii) the mode of debris ejection from the vessel has the dominant effect on the likelihood of creep-rupture failure of the vessel lower head and debris ejection rate from the vessel.

Transient heat transfer in internally heated corium pool

The heat transfer characteristics during different transients are the critical problems in the study of thermal behaviors in volumetrically heated corium pools. In this paper, the COPRA experiments simulating the reactor vessel lower head at 1:1 scale were performed. The nitrate mixture of 20% NaNO3–80% KNO3 in mole fraction was employed as the reactor corium simulant. This study mainly focused on the pool temperature change, heat flux variation and crust thickness evolution during two corium relocation stages and heating power transition stages. Then the transient model was developed and the simulation was validated with COPRA tests and reactor case. The calculation results of primary parameters all agreed well with experimental data during entire tests. Furthermore, the model was applied to reactor situation with prototypical corium and compared with LES results.

In- and ex-vessel coupled analysis for in-vessel retention

In-vessel retention (IVR) in the manner of external reactor vessel cooling (ERVC) is an important severe accident mitigation strategy, which has been applied to some advanced light water reactors, e.g. AP1000. This main assessment method on the effectiveness of IVR is the lumped parameter (LP) model, which has been applied to assess the safety margin of IVR-ERVC for decades. However, this model deals with the safety margin analysis with several isolated processes, without considering the strong coupled relations about the heat transfer processes in the internal and external reactor pressure vessel (RPV). This paper addresses this coupled issue based on modified lumped parameter model using an iterative method, which focuses on the relations between the inner wall surface temperature of the reactor pressure vessel lower head and the heat flux through the RPV inner wall with a modified heat transfer method. The method has been applied to analyze the molten pool behaviour of LIVE-L4 tests. Comparison between the experimental data and the calculation results is performed to validate the accuracy of the coupled analysis method. High attention is paid to some important parameters, e.g. the heat flux through vessel, the inner and outer wall surface temperatures of the RVP lower head and the crust thickness. In addition, the paper analyzes the effectiveness of IVR for AP600 based on the modified method and compares with the UCSB assumed Final Bounding State (FIBS) benchmark calculation results to verify the modified coupled method. This method is applied to conduct the sensitivity analysis for some design parameters of AP600.

Experiment and modeling of jet breakup in fuel-coolant interactions

In light water reactor core melt accidents, the potential risk of energetic molten fuel coolant interaction (FCI, steam explosions) in a partially-flooded reactor cavity has drawn substantial attention. However, when the cavity is completely filled with coolant and the reactor vessel lower head is submerged for the purpose of external reactor vessel cooling, the jet breakup behavior in this case could not only be quite different from the case of partially-flooded reactor cavity, but also affect the risk of fuel coolant interactions. In this study, the jet breakup in liquid-liquid system was experimentally investigated for a fully-flooded and a partially-flooded cavity, and the existing jet breakup models were evaluated against the experimental data. The jet breakup experiment, named as COLDJET, was carried out using the Woods metal whose melting point was 72 °C. The melt jet temperature was lower than the boiling point of coolant so that only the hydrodynamic aspect of jet breakup was investigated. The jet diameter was 50 mm and the coolant depth was 1 m. To investigate the effect of air space of free-falling jet, the experiment was conducted for both fully-flooded and partially-flooded cavities. In the latter case, the free-fall distance in air space was 1 m. A high-speed video camera (typically 4000 fps) was used to visualize the jet breakup and there was an obvious difference in jet breakup behavior between the fully-flooded and the partially-flooded cavities. The jet breakup lengths which were identified in the jet front falling speed were compared with existing models and found that the agreement with the Epstein and Fauske’s model was remarkable. The solidified debris were collected, dried, and sieved to obtain the debris size distribution. It was observed that the average debris size in case of the fully-flooded condition was much smaller than that of the partially-flooded condition at the same jet injection velocity. This is because the entrained air layer between the melt and water acted as like vapor film in boiling condition. The prediction of most probable debris size by the model based on Kelvin-Helmholtz instability showed a good agreement with the experimental data. The jet breakup tests were also simulated using the TRACER-II code, which is a two-dimensional, transient, four-fluid code that can calculate the mixing and propagation of energetic fuel-coolant interactions. The test results were used to improve the jet breakup models of the code as well as to validate the code.

Analysis of Design Extension Conditions in a heavy water research reactor

In this study, the core behavior following the reactor vessel lower head rupture together with the moderator system failure in a tank type heavy water research reactor, IHWRR, is analyzed through the MELCOR code. The focus is on the safety assessment of the reactor core for Design Extension Conditions and overall assessment of safety features of the reactor for any residual risk posed by severe accidents. The inherent features of IHWRR provide a broad spectrum of scenarios where the fuel does not melt, even if primary and moderator cooling are lost. Accordingly, coincident with vessel rupture, three different scenarios are considered for moderator failure: i) loss of moderator forced circulation, calandria tube rupture in ii) the upper and iii) the lower parts of the core. The obtained results showed that the vessel rupture that coincides with the tube rupture in the lower part of the core leads to the oxidation of entire Zr cladding and failure of the calandria tubes without core melt or hydrogen explosion in the containment.

Large eddy simulation on turbulent heat transfer in reactor vessel lower head corium pools

Due to the high-Rayleigh-number natural convection and strong turbulence in the volumetrically heated corium melt pools, it is difficult to capture the heat transfer characteristics in detail. The Large Eddy Simulation (LES) methods are more and more applicable to nuclear industry and have become powerful tools to analyze the complicated turbulent and multi-phase flows. In this paper, the Wall-Modeled LES (WMLES) method was employed for the simulation of three typical corium pool heat transfer experiments: BALI, LIVE and COPRA experiments. The melt pool temperature and heat flux from numerical simulations were in good comparison with experimental data. Then numerical simulation for the transient formation of two-layer corium pool was performed with prototypical corium to obtain the heat transfer data in reactor situation. There was a distinctive gap between the temperatures in two layers. The heat flux showed a decrease at the two-layer interface and a huge jump at the metal layer side. The thin metal layer on the top at the early stage of the two-layer formation will threaten the vessel integrity.

MELCOR severe accident analysis for a natural circulation small modular reactor

A natural circulation small modular reactor (NCSMR) is designed and the severe accident of the NCSMR was analyzed with MELCOR code in this work. The NCSMR is a 330 MWt light water reactor no reactor coolant pumps and no active safety injection systems similar to the design of NuScale reactor. A MELCOR code model was developed for the reactor. The MELCOR model shows consistent results in steady state operation with the design values. The NCSMR severe accident analyzed in this work was induced by the small break loss of coolant accident in the reactor pressure vessel and a crack in the containment pool wall. The decay heat removal system is assumed to fail for the main steam line break. The results indicated that the RPV pressure decreases quickly while the containment pressure increases followed by the opening of reactor vent valves. The core exposes and heat up with the coolant leakage and finally collapsed to the lower plenum, leading to the lower head breach. High temperature core corium quickly broke the relatively thin containment vessel wall, and was discharged to the containment pool floor, leading to the molten corium concrete interaction. The fission products release during the accident process was also analyzed. The severe accident sequence shows that the NCSMR has much longer time duration for the reactor vessel lower head failure to occur compared to conventional commercial nuclear power plant under severe accident conditions. The analysis results help identify appropriate accident management strategies and countermeasures for the potential extreme natural hazard induced severe accidents in NCSMR.

Capabilities of MC3D to investigate the coolability of corium debris beds

A nuclear severe accident progression may lead to the formation of a corium debris bed either in the vessel lower head (in-vessel debris bed) or in the vessel pit (ex-vessel debris bed). For safety analyses it is essential to know if a debris bed is coolable or not, i.e. whether a given water mass flow rate poured into the debris bed – either from its top or from its bottom – will be sufficient to evacuate the residual heat and stop the accident progression. The IRSN code, mostly used for fuel-coolant interaction studies, has been modified with the addition of new friction laws for diphasic flows in porous media. The validation of the code in the case of debris coolability concerns the friction in isothermal configuration in cold and hot situations, the evaluation of critical heat flux and the bottom and top reflooding of debris beds. The results obtained with MC3D are in good agreement with the experimental data and are estimated satisfactory regarding to the nuclear safety issues.

A simplified geometrical model for transient corium propagation in core for LWR with heavy reflector

In the context of the simulation of the Severe Accidents (SA) in Light Water Reactors (LWR), we are interested on the in-core corium pool propagation transient in order to evaluate the corium relocation in the vessel lower head. The goal is to characterize the corium and debris flows from the core to accurately evaluate the corium pool propagation transient in the lower head and so the associated risk of vessel failure. In the case of LWR with heavy reflector, to evaluate the corium relocation into the lower head, we have to study the risk associated with focusing effect and the possibility to stabilize laterally the corium in core with a flooded down-comer. It is necessary to characterize the core degradation and the stratification of the corium pool that is formed in core. We assume that the core degradation until the corium pool formation and the corium pool propagation could be modeled separately. In this document, we present a simplified geometrical model (0D model) for the in-core corium propagation transient. A degraded core with a formed corium pool is used as an initial state. This state can be obtained from a simulation computed with an integral code. This model does not use a grid for the core as integral codes do. Geometrical shapes and 0D models are associated with the corium pool and the other components of the degraded core (debris, heavy reflector, core plate…). During the transient, these shapes evolve taking into account the thermal and stratification behavior of the corium pool and the melting of the core surrounding components. Some results corresponding to the corium pool propagation in core transients obtained with this model on a LWR with a heavy reflector are given and compared to grid approach of the integral codes MAAP4.

Core Melt Composition at Fukushima Daiichi: Results of Transient Simulations with ASTEC

The Accident Source Term Evaluation Code (ASTEC) is used to perform numerical simulations of the accidents at the Fukushima Daiichi nuclear power station in the frame of the Organisation for Economic Co-operation and Development/Nuclear Energy Agency Benchmark Study of the Accident at the Fukushima Daiichi Nuclear Power Station (BSAF) project. At present, simulations are available for Units 1, 2, and 3 of Fukushima Daiichi and for 6 days from the earthquake. A clear lesson from phase 1 of the project was that the uncertainties in the functioning of the safety systems and in accident progression are still large and there are many explanations for the measured thermohydraulic behavior. Rather than focusing on the thermohydraulic key parameters for which comparisons with measurements are available, this paper will address melt composition computation results that may provide insights relevant for the decommissioning process.When molten corium relocates from the core down to the vessel lower head, the melt jets interact with water and may be totally or partially fragmented depending on the level of water. A U-Zr-O-Fe molten pool may form in the lower head, and because of chemical reactions, separation between nonmiscible metallic and oxide phases may occur. The models implemented in ASTEC enable the simulation of these phenomena. Up to five different axisymmetric corium layers in the vessel bottom head can be formed, which are, from bottom to top, a debris layer, a heavy metallic layer, an oxide layer, a light metallic layer, and another debris layer. An important process is the UO2 fuel reduction to metallic uranium by nonoxidized zirconium, which results in uranium transport to the dense metallic layer as demonstrated in the MAterial SCAling (MASCA) program.Because of the large consensus on the accident progression of Fukushima Daiichi Unit 1, in this paper we present complex melt compositions before vessel failure for the current best-estimate cases for Unit 1. We do not present similar work performed for Units 2 and 3.It should be underlined that in the case of vessel bottom failure, a part of this complex melt will relocate to the pedestal and molten core–concrete interaction will take place enhancing other complex physical phenomena with possible large consequences on the melt chemical composition and behavior.

Efficacy of endoscopic transpapillary drainage for patients with pancreatic duct disruption

In vessel retention (IVR) is one of the key strategies for many advanced LWR designs to mitigate postulated severe accidents. The success of IVR substantially relies on external reactor vessel cooling (ERVC) by which the decay heat is removed from the melt core in the reactor vessel lower head. The main challenge of IVR is to provide an adequate safety margin of ERVC against critical heat flux (CHF) of subcooled flow boiling in the reactor lower head flow channel. Due to uncertainties in corium melt pool configuration, large CHF margin of ERVC is usually required by regulatory authorities to demonstrate reliability of severe accident mitigation methods. Various CHF enhancement designs have been proposed and studied in literature. In this paper, an experimental study of hypervapotron structure as a novel design to improve CHF performance of ERVC is conducted. Hypervapotron is chosen as one of the potential engineering options for International Thermonuclear Experimental Reactor (ITER) program as a divertor structure to remove highly intense heat from fusion chamber. This study is to conduct CHF experiments at typical PWR ERVC working conditions. The CHF experiments are performed in a 30 mm by 61 mm rectangular flow channel with a 200 mm long heated surface along the flow direction. Both smooth and hypervapotron surface are tested at various inclination angles of the test section to simulate various positions of the reactor lower head. The hypervapotron is found to have a 40–60% CHF improvement compared with the smooth surface. The high speed visualization indicates that hypervapotron is able to effectively remove big vapor mushrooms on the heating surface.