High-Temperature Characterization of Melted Nuclear Core Materials: Investigating Corium Properties Through the Case Studies of In-Vessel and Ex-Vessel Retention (original) (raw)
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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...
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.
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 7th 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 ...
Modeling of corium melt cooling during severe accidents at the nuclear power plants
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Behavior of the corium pool in the lower head is still a critical issue in understanding of Pressurized Water Reactor (PWR) core meltdown accidents. One of the key parameter for assessing the vessel mechanical strength is the resulting heat flux at the pool-vessel interface. A number of studies [1]-[3] have already been performed to pursue the understanding of a severe accident with core melting, its course, major critical phases and timing and the influence of these processes on the accident progression. Uncertainties in modeling these phenomena and in the application to reactor scale will undoubtedly persist. These include e.g. formation and growth of the in-core melt pool, relocation of molten material after the failure of the surrounding crust, characteristics of corium arrival in residual water in the lower head, corium stratifications in the lower head after the debris re-melting [4]. These phenomena have a strong impact on a potential termination of a severe accident. The main objective of the LIVE program [5] at Karlsruhe Institute of Technology (KIT) is to study the core melt phenomena both experimentally in large-scale 3D geometry and in supporting separate-effects tests, and analytically using CFD codes in order to provide a reasonable estimate of the remaining uncertainty band under the aspect of safety assessment. Within the LIVE experimental program several tests have been performed with water and with non-eutectic melts (mixture of KNO 3 and NaNO 3) as simulant fluids. The results of these experiments, performed in nearly adiabatic and in isothermal conditions, allow a direct comparison with findings obtained earlier in other experimental programs (SIMECO, ACOPO, BALI, etc.) and will be used for the assessment of the correlations derived for the molten pool behavior. The information obtained from the LIVE experiments includes heat flux distribution through the reactor pressure vessel wall in transient and steady state conditions, crust growth velocity and dependence of the crust formation on the heat flux distribution through the vessel wall. Supporting posttest analysis contributes to characterization of solidification processes of binary non-eutectic melts. Complimentary to other international programs with real corium melts, the results of the LIVE activities provide data for a better understanding of incore corium pool behavior. The experimental results are being used for development of mechanistic models to describe the incore molten pool behavior and their implementation in the severe accident codes like ASTEC. The paper summarizes the objectives of the LIVE program and presents the main results obtained in the LIVE experiments up to now.
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Melting of a reactor core is a multilevel process that can proceed along more than one path. Investigation of this process should be oriented toward studying the properties and quantitative ratios of phases forming conglomerates and melts in the core space. Data on the properties of the phases and their quantitative ratio can form a base for calculating the properties of conglomerates and melts over a wide range of concentrations. The correctness of estimating the properties of complicated alloys on the basis of a knowledge of the properties of individual components and the adequacy of the model can be checked experimentally by comparing the measured and computed properties of reference alloys of a complicated composition.
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Nuclear Engineering and Design, 2018
The paper presents some results of the ISTC (International Science and Technology Center)-financed project 'Investigation of Corium Melt Interaction with NPP Reactor Vessel Steel' (METCOR). In the METCOR experiments the metallic phase of a two-liquid system was produced by the interaction between hot suboxidized corium and cooled VVER vessel steel, with the steel being corroded. Models of corrosion mechanisms in the considered conditions are used to systematize data on the limiting temperature of corrosion/(dissolution) of the vessel steel. A considerable influence of thermal gradient conditions is shown, which has to be taken into account in the analysis of molten pool behaviour.