High pressure simulation experiment on corium dispersion in direct containment heating (original) (raw)
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Experimental simulation of corium dispersion phenomena in direct containment heating
Nuclear Engineering and Design, 1996
In a direct containment heating (DCH) accident scenario, the degree of corium dispersion is one of the most significant factors responsible for the reactor containment heating and pressurization. To study the mechanisms of the corium dispersion phenomenon, a DCH separate effect test facility of 1:10 linear scale for Zion PWR geometry is constructed. Experiments are carried out with air water and air-woods metal simulating steam and molten core materials. The physical process of corium dispersion is studied in detail through various instruments, as well as with flow visualization at several locations. The accident transient begins with the liquid jet discharge at the bottom of the reactor pressure vessel. Once the jet impinges on the cavity bottom floor, it immediately spreads out and moves rapidly to the cavity exit as a film flow. Part of the discharged liquid flows out of the cavity before gas blowdown, and the rest is subjected to the entrainment process due to the high speed gas stream. The liquid film and droplet flows from the reactor cavity will then experience subcompartment trapping and re-entrainment. Consequently, the dispersed liquid droplets that follow the gas stream are transported into the containment atmosphere, resulting in containment heating and pressurization in the prototypic condition. Comprehensive measurements are obtained in this study, including the liquid jet velocity, liquid film thickness and velocity transients in the test cavity, gas velocity and velocity profile in the cavity, droplet size distribution and entrainment rate, and the fraction of dispersed liquid in the containment building. These data are of great importance for better understanding of the corium dispersion mechanisms.
Corium Dispersion and Direct Containment Heating Experiments at Low System Pressure
Experiments in a reduced scale were performed with an iron-alumina melt, steam and a prototypic atmosphere in the containment, to investigate the fluid-dynamic, thermal and chemical processes dur-ing melt ejection out of a breach in the lower head of a PWR pressure vessel at pressures below 2 MPa. A cavity geometry with a direct path into the containment and one with a closed reactor pit, where the only flow path out of the pit is along the main cooling lines leading into reactor rooms, were investigated. Also, an experiment with nitrogen driven melt is compared to one with steam driven melt. With a closed reactor pit, there will be a considerable melt ejection into the pump and steam generator rooms, but almost nothing into the open space of the containment. The pressure increase will stay moderate and well below the design pressure of most containments. The comparison of two tests with and without steam, showed the strong effect of hydrogen production and combustion on both, the m...
Low-Pressure Corium Dispersion Experiments with Simulant Fluids in a Scaled Annular Cavity
Experiments were performed in a scaled annular cavity design, to investigate melt dispersal from the reactor pit when the reactor pressure vessel (RPV) lower head fails at low system pressure of less than 2 MPa. The fluid dynamics of the dispersion process was studied using model fluids, water, or bismuth alloy instead of corium, and nitrogen or helium instead of steam. The effects of different breach sizes and locations and different failure pressures on the dispersion were studied, specifically by testing central holes, lateral holes, horizontal rips, and complete unzipping of the bottom head.With holes at the base of the bottom head, the most important parameters governing the dispersion of melt are the hole size and the burst pressure. The fraction dispersed into the reactor compartments increases with larger holes and higher pressures. Values up to 76% have been found for both melt simulant liquids, water, and metal. With lateral breaches the liquid height in the lower head rel...
Experiments were performed in a scaled annular cavity design, to investigate melt dispersal from the reactor pit when the reactor pressure vessel lower head fails at low system pressure of less than 2 MPa. In the part of the experimental program presented in this paper, the fluid dynamics of the dispersion process was studied using model fluids, water or bismuth alloy instead of corium, and nitrogen or helium instead of steam. The effects of different breach sizes and locations, and different failure pressures on the dispersion were studied, specifically by testing central holes, lateral holes, horizontal rips, and complete ripping of the bottom head. The experiments have shown, that lateral failures lead to smaller melt dispersal out of the reactor pit than failures in the central part of the lower head. With holes at the base of the bottom head, the most important parameters governing the dispersion of melt are the hole size and the burst pressure. The fraction dispersed into the ...
Corium behavior and steam explosion risks: A review of experiments
Annals of Nuclear Energy, 2018
After the Three Mile Island accident, numerous studies on the severe nuclear accident have been conducted. In a degraded core accident, the high-temperature melt corium may drop into the lowtemperature coolant, which is called the fuel coolant interaction (FCI). Due to the velocity difference between the melt jet and the coolant as well as the violent film boiling around the melt, the melted corium may fragment into small particles. With the increase of the contact area between the melted corium and the coolant, plenty of coolant steam is produced. The timescale for heat transfer is shorter than that for pressure relief, resulting in the formation of shock waves and/or the production of missiles at a later time during the expansion of coolant steam explosion. During a severe reactor accident scenario, steam explosion is an important risk, even though its probability to occur is pretty low, since it could lead to large releases of radioactive material, and destroy the reactor vessel and surrounding structures. This study provides a comprehensive review of vapor explosion experiments, especially the most recent ones. In this review, small-to intermediate-scale experiments related to premixing, triggering and propagation stages are first reviewed and summarized in tables. Then intermediate-to large-scale experiments using prototypic melt are reviewed and summarized. The recent OECD/SERENA2 project including KROTOS and TROI facilities' work is also discussed. The studies on steam explosion are vital for reactor severe accident management and will lead to improved reactor safety.
Study of the processes of corium-melt retention in the reactor pressure vessel (INVECOR)
Integral large-scale vessel retention experiments have been performed using up to 60 kg of prototypic corium melt discharged from the electric melting furnace at a height of 1,7 m into a model RPV (Reactor Pressure Vessel-40cm dia. x 60cm depth) with plasmatrons for decay heating of corium for 1-2 hours. Specific power release in corium was 6-9 W.cm -3 and the maximum temperature of the RPV wall was up to 1400°C. The following has been achieved during the project: 1) Protective coatings on the graphite crucibles and the plasmatron graphite nozzles have been further developed. Numerous trials were carried out to improve the decay heat simulation of corium. 2) Calculations of the corium pool (heating efficiency, thermal fluxes and temperature distributions) were performed with specific tests for validation of the models. 3) 4 large-scale experiments with the model RPV using a molten oxidic corium and oxidic-metallic corium were conducted. 4) Extensive post-test analysis of corium samp...
Frontiers in Energy Research
During a severe accident in a nuclear reactor, the molten core—or corium—may be relocated into the reactor vessel’s lower plenum in case of core support plate failure. The severe accident management strategy for In-Vessel Retention—or IVR—consists in stabilizing the corium within the reactor pressure vessel by external cooling of the vessel’s lower head. If now, the vessel fails due to excessive thermal loading on its walls, the Ex-Vessel Retention—or EVR—strategy is adopted. In this case, the core melt stabilization can be achieved by effective corium spreading, either in the reactor vessel cavity or in a dedicated “core-catcher”, and cooling by water. The success of both strategies highly depends on the corium behavior at high temperatures, conditioning vessel’s integrity for IVR, and promotion for the spreading of the EVR. This involves a variety of fundamental mechanisms closely related to heat and mass transfer regimes prevailing at the system scale, which requires further anal...
2010
This paper discusses an approach for application of the computational fluid dynamics (CFD) method to support development and validation of computationally effective methods for safety analysis, on the example of molten corium coolability in a BWR lower head. The approach consists of five steps designed to ensure physical soundness of the effective method simulation results: (i) analysis and decomposition of a severe accident problem into a set of separate-effect phenomena, (ii) validation of the CFD models on relevant separate-effect experiments for the reactor prototypical ranges of governing parameters, (iii) development of effective models and closures on the base of physical insights gained from relevant experiments and CFD simulations, (iv) using data from the integral experiments and CFD simulations performed under reactor prototypic conditions for validation of the effective model with quantification of uncertainty in the prediction results and (v) application of the computationally effective model to simulate and analyze the severe accident transient under consideration, including sensitivity and uncertainty analysis. Implementation of the approach is illustrated on a so-called effective convectivity model for simulation of turbulent natural convection heat transfer and phase changes in a decay-heated corium pool. It is shown that detailed information obtained from the CFD simulations are instrumental to ensure the effective models capture safety-significant local phenomena, e.g. the enhanced downward heat flux in the vicinity of a cooled control rod guide tube.
On the basis of reasonable core meltdown conditions that can be postulated for GenIV sodium fast reactors, during a severe accident, good safety margins can be achieved for corium confinement and cooling inside the reactor vessel, by the use, in the lower plenum, of a core catcher. Such a device has to be designed to withstand to extreme thermal-mechanical conditions that rise as consequence of the large mechanical energy release and high temperature of molten corium. In the frame of the activities carried out within the CP-ESFR Project of the 7 th Framework Programme Euratom, and considering the postulated accident conditions assumed for a reference 1500 MWe pool-type sodium fast reactor, the present work provides a preliminary analysis of the thermal response of a possible core catcher placed within the vessel. The dynamic thermal behaviour of the corium-structure-coolant system is analyzed with the computer code CORIUM-2D, an original simulation tool developed by RSE with the aim to assess the thermal interaction among corium, structures and coolant under severe accident conditions in both LWRs and LMFBRs. Temperatures reached by the core catcher, vessel and safety vessel, at the end of thermal transients, have been compared with safety criteria assumed for the demonstration of the corium coolability and long term integrity of these components. The results show that the steady-state coolable configuration of core debris and the structural integrity of main containment structures can be reached in a limited number of partial core meltdown situations.