Coolability Of Corium Research Articles

Simulation of melt spreading over dry substrates with the moving particle Semi-implicit method

In a severe accident scenario of a light water reactor, the spreading of molten core materials (corium) on the floor of the reactor cavity may occur after the failure of the reactor pressure vessel (RPV), resulting in the so-called melt spreading process which is important to assess the melt progression towards containment structures (e.g., the liner) and the potential for ex-vessel coolability of corium. The present study is concerned with the numerical simulation of melt spreading on a dry floor using the Moving Particle Semi-implicit (MPS) method which is mesh free and suitable for such a free-surface flow. Multi-physics models are implemented in the MPS method to enable modeling of physical phenomena occurring in melt spreading process. The emphasis is to develop a crust formation model which is not only appropriate for capturing crust formation at the upper surface but also capable of reproducing the shape of the spread. A genetic algorithm (GA) is also employed to optimize the initial distance between particles. The multi-physics models and GA are coded into a computer program using C programming language. The MPS code is then used to simulate the small-scale pre-test and S3E 3MDS-Ox-2 melt spreading experiment conducted at KTH. The simulation results show the capability of the MPS code predicting melt leading-edge progression.

Modeling crust fracture and water ingression through crust during top-flooding strategy for corium cooling

Abstract During an LWR severe accident, top flooding strategy is usually used in order to achieve the coolability of corium. Water ingression phenomenon, where water percolates through cracks in the crust, is identified as one of the phenomena that enhance the heat removal during cooling of the corium via top flooding. In this study, a model for the water ingression phenomenon is suggested. The model assumes that the morphology of fractured crust is determined during the period before onset of water ingression. Especially, the present model predicts the morphology of the crust using two thermal-stress based criteria: a tensile-strength criterion and a toughness criterion for the fracture. In addition, unlike previous models that assume heat removal is limited hydraulically to the dryout heat flux, heat flux during water ingression is estimated considering the phenomenon where quenching is limited thermally by large thermal energy of crust. When the current model was used to estimate the heat removal rate during the quenching of the fractured crust, it showed 26.5% of root-mean-square error (RMSE) for data (Lomperski and Farmer, 2007), which was improved from 39.3% RMSE of the existing hydraulic-limit based models (Lomperski and Farmer, 2007). In addition, comparing removed corium energies calculated from the present model with one calculated without considering water ingression, we found that water ingression can remove three-times larger energy when the concrete content in the corium is lower than 5 wt%.

In-Calandria Retention of Corium in PHWR: Experimental Investigation and Remaining Issues

After the Fukushima accident, the public has expressed concern regarding the safety of nuclear power plants. This accident has strengthened the necessity for further improvement of safety in the design of existing and future nuclear power plants. Pressurized heavy water reactors (PHWRs) have a high level of defense-in-depth (DiD) philosophy to achieve the safety goal. It is necessary for designers to demonstrate the capability of decay heat removal and integrity of containment in a PHWR reactor for prolonged station blackout to avoid any release of radioactivity in public domain. As the design of PHWRs is distinct, its calandria vessel (CV) and vault cooling water offer passive heat sinks for such accident scenarios and submerged calandria vessel offers inherent in-calandria retention (ICR) features. Study shows that, in case of severe accident in PHWR, ICR is the only option to contain the corium inside the calandria vessel by cooling it from outside using the calandria vault water to avoid the release of radioactivity to public domain. There are critical issues on ICR of corium that have to be resolved for successful demonstration of ICR strategy and regulatory acceptance. This paper tries to investigate some of the critical issues of ICR of corium. The present study focuses on experimental investigation of the coolability of molten corium with and without simulated decay heat and thermal behavior of calandria vessel performed in scaled facilities of an Indian PHWR.

Experimental investigation of heat transfer during severe accident of a Pressurized Heavy Water Reactor with simulated decay heat generation in molten pool inside calandria vessel

The present study focuses on experimental investigation in a scaled facility of an Indian PHWR to investigate the coolability of molten corium with simulated decay heat in the simulated calandria vessel. Molten borosilicate glass was used as the simulant due to its comparable heat transfer characteristics similar to prototypic material. About 60kg of the molten material was poured into the test section at about 1200°C. Decay heat in the melt pool was simulated using four high watt heaters cartridges, each having 9.2kW. The temperature distributions inside the molten pool, across the vessel wall thickness and vault water were measured. Experimental results obtained are compared with the results obtained previously for no decay heat case. The results indicated that presence of decay heat seriously affects the coolability behaviour and formation of crust in the melt pool. The location and magnitude of maximum heat flux and surface temperature of the vessel also are affected in the presence of decay heat.

Quenching Behaviour of Top Flooded Molten Pool

During a severe accident in a nuclear reactor, the core can melt and the melt can be relocated into the lower head forming a melt pool. If the vessel fails, the molten corium can be relocated in the containment cavity forming a melt pool. Such a pool or bed, if not quenched in time, may interact with the concrete basement of the cavity causing its ablation resulting in generation of non-condensable gases and water vapor, which poses a threat of containment pressurization, explosion and ground contamination. In order to devise the strategy to retard the progression of severe accidents in stipulated time, understanding of the corium coolability is very much essential. Coolability of molten corium in such ex-vessel condition is limited depending on depth of water ingression. But the question arises about to what extent the water will ingress? Thermal and physical properties of corium change drastically with increase in concrete percentage which strongly affects the coolability. The phenomenon of corium coolability is not fully understood owing to its complexity involving multi phase multi component heat and mass transfer. In order to better understand the phenomenon, experiments at present are the only answer; since modeling the complex phenomenon is still difficult. In order to gain some insights into melt coolability, the authors have carried out an experiment on quenching behavior of top flooded molten pool. A simulant material (sodium borosilicate glass) of about 25 liters at temperature 1200 oC was poured into the test section and was flooded from top with water. The transient temperature of the molten pool was measured. The experiment highlighted that, under adiabatic conditions, water ingression occurred only upto 10 mm depth, below which a stable solid crust was formed which limited the heat transfer. Also, no gap between crust and vessel was formed.

A simple model to understand physics of melt coolability under bottom flooding

Following a severe accident in a nuclear reactor, stabilization and termination of molten pool/debris is necessary to ensure safety of the reactor. In this regard, ex-vessel corium coolability still remains an unresolved issue in spite of several efforts being taken towards its understanding. Extensive research has been carried out in the field of corium coolability phenomenon, but the outcome is far from establishing a certainty of complete coolability of melt under all possible scenarios. Major thrust so far was on top flooding condition, which often results in uncoolable condition due to inability of water ingression deep into the solid crust. The research on coolability under bottom flooding condition started long back, but has gained lot of momentum recently. However, mainly the focus is on development of engineering system to enhance the coolability using bottom flooding. The understanding of the phenomenon is not yet totally brought out. In this paper, we have postulated the mechanism of melt coolability under bottom flooding. We have developed a simplified mechanistic model for prediction of coolability of molten corium under bottom flooding condition based on the postulations. The model has been validated with the experimental data from literature. Comparison of top flooding vis-à-vis bottom flooding has also been presented.

Interfacial drag of two-phase flow in porous media

Abstract In this paper, the influence of the interfacial drag on the pressure loss of combined liquid and vapour flow through particulate porous media is investigated. Motivation for this is the coolability of fragmented corium which may be expected during a severe accident in a nuclear power plant. Cooling water is evaporated due to the particles decay heat. To reach coolability, the outflowing steam has to be replaced by inflowing water. The pressure loss for the resulting two-phase flow through the debris is determined by the friction forces at the particles as well as at the interfacial drag between the fluids. In return, the resulting pressure field influences the coolant inflow through the boundaries and thus the coolability of the fragments. Starting from commonly used friction correlations, it is shown that only those with explicit consideration of the interfacial drag can represent the pressure loss. Enhancements for the formulation of the interfacial drag are proposed. This modified model fits experimental data of isothermal air/water experiments as well as data of boiling particle beds.

Boiling experiments for the validation of dryout models used in reactor safety

The coolability of fragmented corium is a major issue in reactor safety. Since the long-term coolability of such particle beds is limited by the availability of coolant inside the bed and not by heat transfer limitations from the particles to the coolant, the pressure field inside the debris has a strong effect on the cooling potential in multi-dimensional cases as expected in severe accidents in light water reactors (LWR). Therefore, the determination of the pressure field for two-phase flows in porous media is one central point of interest. In this context simulation models and in particular dryout models were developed for reactor safety analyses which have to be validated by reliable experimental data. Therefore, basic experimental investigations have been carried out with inductively heated steel balls of 6 or 3 mm diameter to provide a database for the validation and modification of the friction laws included in these dryout models. The performed boiling and dryout experiments show clearly that models without the explicit consideration of the interfacial drag cannot predict the pressure distribution inside a boiling particle bed, not even qualitatively. Against it, models with an explicit consideration of the interfacial drag can describe the distribution of pressure inside a boiling particle bed.