Sensitivity Analysis of Ex-Vessel Corium Coolability Models in MAAP5 Code for the Prediction of Molten Corium–Concrete Interaction after a Severe Accident Scenario (original) (raw)
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Nuclear Engineering and Technology
In the present study, the cavity module of the MELCOR code is used for the simulation of molten corium concrete interaction (MCCI) during the late phase of postulated large break loss of coolant (LB-LOCA) accident in the APR1400 reactor design. Using the molten corium composition data from previous MELCOR Simulation of APR1400 under LB-LOCA accident, the ex-vessel phases of the accident sequences with long-term MCCI are recalculated with stand-alone cavity package of the MELCOR code to investigate the impact of water ingression and melt eruption models which were hitherto absent in MELCOR code. Significant changes in the MCCI behaviors in terms of the heat transfer rates, amount of gases released, and maximum cavity ablation depths are observed and reported in this study. Most especially, the incorporation of these models in the new release of MELCOR code has led to the reduction of the maximum ablation depth in radial and axial directions by ~38% and ~32%, espectively. These impacts are substantial enough to change the conclusions earlier reached by researchers who had used the older versions of the MELCOR code for their studies. and it could also impact the estimated cost of the severe accident mitigation system in the APR1400 reactor.
Modeling of corium melt cooling during severe accidents at the nuclear power plants
This paper is devoted to an analysis of the problem of a corium melt interaction with the water and low-melting temperature blocks in the passive protection systems against severe accidents at the nuclear power plants (NPP), which is of high importance for a substantiation of a nuclear power safety, for building and successful operating of the passive protection systems. In the third-generation reactors the passive protection systems against severe accidents at the NPP are mandatory, therefore the topic of this paper is of importance for the nuclear power safety. A few such systems have been considered, which are in different stage of completeness.
2015
The MELCOR and CORQUENCH codes are used for simulations of the late phase of a postulated total SBO at KWU KONVOI PWR reactor. The base line SBO sequence without any mitigation and some of its variants are simulated in detail with MELCOR, selected cases up to 10 days of the accident progression. The ex-vessel phases of these scenarios, with long-term molten core-concrete interactions (MCCI), are recalculated also with the specialized code CORQUENCH. The AM counter measures after the reactor pressure vessel failure, typically flooding of the melt by water from the top at different times, are taken into account. The incentive for the work was to investigate the retention of corium materials, which were subject to MCCI, for this particular type of containment. At the same time, the detailed analysis is performed of the effect of different models employed for the description of heat transfer at the melt-concrete interface (ablation modeling) and of the effect of the "coolability" models. The results of the calculations indicate that this type of the KONVOI containment is relatively resistant to challenges by the late-phase accident progression. The incorporation of the CORQUENCH coolability models into MELCOR/CORCON, where it is missing, should be of primary interest. The model of the heat conduction into concrete behind the ablation front could also be useful. The overall impact of this modeling option-which is available in CORQUENCH and not available in MELCOR-on the maximum ablation depth seems to be relatively high under conditions pertaining to our accident simulations.
Energies
The characterization of molten corium–concrete interaction (MCCI) has increasingly become a cause of concern because, in the case of a severe nuclear accident, the core could meltdown and release radiation into the environment. The objective of this study was to determine the thermochemical impact of metal content in the corium and analyze the effect of corium metal content on ablation depth, corium temperature, its viscosity and surface heat flux, and production of hydrogen, carbon monoxide, and carbon dioxide. The governing heat transfer equations were solved while considering the various thermochemical reactions in the existing numerical code in a comprehensive way. The developed MCCI model in CORQUENCH was validated against the data available in the literature. Our findings showed that the composition of corium, especially its metal content, has a noticeable effect on mitigating or aggravating the ablation depth and nuclear reactor integrity. We observed that during molten coriu...
Fire Technology
This paper presents some experimental and numerical results on the study of the hygrothermal behavior of the High Performance Concrete used to build the inner wall of the nuclear power station of CIVAUX 2 (in France). In the laboratory, accidental situations (beyond design) have been considered. The scenario, designated as Severe Accident (SA), consisted of a rise from ambient conditions to 200C, and a steam pressure of 1.3 MPa in 24 hours, maintained after several hours. Those conditions were applied on one face of a specimen, the other one being in ambient temperature. A cylindrical specimen with 1.3 m of thickness was used. Thermocouples, pressure taps, and moisture meters were implanted in the concrete at the moment of casting. They gave local information; most of them are distributed in the first 0.30 m. The typical experimental results of the evolution of the temperature, the pressure and water content as a function of time, are shown for the testing conditions. The thickness ...
Progress in Nuclear Energy, 2022
Molten corium, a mixture of molten nuclear fuel, cladding, thermo-hydraulic and structural elements, can originate in a nuclear plant accident after a reactor core meltdown. This un-cooled corium could penetrate through the reactor pressure vessel and cause concrete ablation via basement melt-through, a process known as Molten Corium Concrete Interaction (MCCI). The MCCI analysis because of its complex nature is still uncertain and needs thorough investigation of various parameters. In this study the use of CORQUENCH simulator is presented to model the molten corium, composition of concrete and heat transfer along with related chemical reactions. Using this modeling technique, the chemical reaction capabilities of CORQUENCH is successfully utilized enabling the modeling of interaction between molten corium and concrete. The developed model is validated against experimental data at PWR and BWR conditions. The results showed that the temperature of corium, composition of concrete and water injection time have a pronounced effect on mitigating ablation and reactor integrity in case of a nuclear accident. In addition, the composition of concrete was found to be the main controlling factor to mitigate ablation. An alternative to concrete is to utilize igneous rock (pyrolite) and this approach could lead to comparatively very low rates of ablation due to its high thermal resistant properties. Furthermore, the injection of water (as a cooling agent) into the reactor cavity should also be optimized to enhance corium quenching to avoid ablation via basement melt-through. The concrete ablation mechanisms during MCCI are very case-dependent on the concrete solidus, liquidus and ablation temperatures, respectively.
2015
Reactor core degradation and in-vessel and ex-vessel corium behavior have been major research topics for the last three decades to which IRSN strongly contributed by the coordination of, or by the contribution to, large R&D programs and through the development and validation of the severe accident ASTEC code. In the last years, the balance of research efforts tipped on analyses of pro and cons and assessments of mitigation measures. The outcomes of risk significance analysis (including fuel-coolant interaction (FCI), hydrogen combustion and molten core concrete interaction (MCCI) risks) performed in France as well as the retained orientations for corium behavior research are described. The focus lies nowadays in (1) invessel melt retention (IVMR) strategies for future reactor concepts but also related to the need to establish the reliability of such strategies when implemented in existing reactors (2) in-containment corium cooling for existing reactors. The paper summarizes main ach...
2015
Molten Corium-Concrete Interaction (MCCI) and corium coolability (heat transfer between corium and overlying water pool) are modeled as part of the containment model in the MAAP 5 code. MAAP5.03+ (an alpha version of what will eventually become MAAP5.04) has added bulk cooling as one of the corium coolability mechanisms. There were two mechanisms modeled in previous versions of the code: water ingression and melt eruption. This paper discusses the features of the MCCI and corium coolablity models in MAAP5, and presents the benchmark results of comparisons of MAAP5.03+ against two tests in the OECD/CCI test series: CCI-2 and CCI-3. It discusses the comparisons with and without the bulk cooling model to demonstrate the performance of the models.
Sensitivity study on severe accident core melt progression for advanced PWR using MELCOR code
Nuclear Engineering and Design, 2014
A LBLOCA scenario for an advanced pressurized water reactor, call APR1400, developed in Korea is analyzed in order to obtain an overall insight into a severe accident progression from an initiating event to the reactor vessel failure in detail by using the MELCOR computer code Versions 1.8.5 and 1.8.6. The present results (the amount of molten corium and vessel failure timing) would be used as input for the establishment of severe accident management strategies or for the design of a core catcher for the APR1400. The MELCOR results showed that the lower head instrumentation tube penetration failure model and internal structure in the reactor vessel had influence on the amount of corium ejected and the timing of reactor vessel failure.
Modelling of Severe Accident and In-Vessel Melt Retention Possibilities in BWR Type Reactor
Science and Technology of Nuclear Installations, 2018
One of the severe accident management strategies for nuclear reactors is the melted corium retention inside the reactor pressure vessel. The work presented in this article investigates the application of in-vessel retention (IVR) severe accident management strategy in a BWR reactor. The investigations were performed assuming a scenario with the large break LOCA without injection of cooling water. A computer code RELAP/SCDAPSIM MOD 3.4 was used for the numerical simulation of the accident. Using a model of the entire reactor, a full accident sequence from the large break to core uncover and heat-up as well as corium relocation to the lower head is presented. The ex-vessel cooling was modelled in order to evaluate the applicability of RELAP/SCDAPSIM code for predicting the heat fluxes and reactor pressure vessel wall temperatures. The results of different ex-vessel heat transfer modes were compared and it was concluded that the implemented heat transfer correlations of COUPLE module i...