Corium Pool Research Articles

New set of convective heat transfer coefficients established for pools and validated against CLARA experiments for application to corium pools

During an hypothetical Pressurized Water Reactor (PWR) or Boiling Water Reactor (BWR) severe accident with core meltdown and vessel failure, corium would fall directly on the concrete reactor pit basemat if no water is present. The high temperature of the corium pool maintained by the residual power would lead to the erosion of the concrete walls and basemat of this reactor pit. The thermal decomposition of concrete will lead to the release of a significant amount of gases that will modify the corium pool thermal hydraulics. In particular, it will affect heat transfers between the corium pool and the concrete which determine the reactor pit ablation kinetics.A new set of convective heat transfer coefficients in a pool with different lateral and horizontal superficial gas velocities is modeled and validated against the recent CLARA experimental program. 155 tests of this program, in two size configurations and a high range of investigated viscosity, have been used to validate the model. Then, a method to define different lateral and horizontal superficial gas velocities in a 0D code is proposed together with a discussion about the possible viscosity in the reactor case when the pool is semi-solid. This model is going to be implemented in the 0D ASTEC/MEDICIS code in order to determine the impact of the convective heat transfer in the concrete ablation by corium.

Applications of ASTEC integral code on a generic CANDU 6

In case of a hypothetical severe accident in a nuclear power plant, the corium consisting of the molten reactor core and internal structures may flow onto the concrete floor of containment building. This would cause an interaction between the molten corium and the concrete (MCCI), in which the heat transfer from the hot melt to the concrete would cause the decomposition and the ablation of the concrete. The potential hazard of this interaction is the loss of integrity of the containment building and the release of fission products into the environment due to the possibility of a concrete foundation melt-through or containment over-pressurization by the gases produced from the decomposition of the concrete or by the inflammation of combustible gases. In the safety assessment of nuclear power plants, it is necessary to know the consequences of such a phenomenon.The paper presents an example of application of the ASTECv2 code to a generic CANDU6 reactor. This concerns the thermal-hydraulic behaviour of the containment during molten core–concrete interaction in the reactor vault. The calculations were carried out with the help of the MEDICIS MCCI module and the CPA containment module of ASTEC code coupled through a specific prediction–correction method, which consists in describing the heat exchanges with the vault walls and partially absorbent gases. Moreover, the heat conduction inside the vault walls is described. Two cases are presented in this paper taking into account two different heat transfer models at the pool/concrete interface and siliceous concrete. The corium pool configuration corresponds to a homogeneous configuration with a detailed description of the upper crust.

Evaluation of heat-flux distribution at the inner and outer reactor vessel walls under the in-vessel retention through external reactor vessel cooling condition

Background: A numerical simulation was carried out to investigate the difference between internal and external heat-flux distributions at the reactor vessel wall under in-vessel retention through external reactor vessel cooling (IVR-ERVC). Methods: Total loss of feed water, station blackout, and large break loss of coolant accidents were selected as the severe accident scenarios, and a transient analysis using the elementbirth-and-death technique was conducted to reflect the vessel erosion (vessel wall thickness change) effect. Results: It was found that the maximum heat flux at the focusing region was decreased at least 10% when considering the two-dimensional heat conduction at the reactor vessel wall. Conclusion: The results show that a higher thermal margin for the IVR-ERVC strategy can be achieved in the focusing region. In addition, sensitivity studies revealed that the heat flux and reactor vessel thickness are dominantly affected by the molten corium pool formation according to the accident scenario.

Numerical Modelling of Melt Behaviour in the Lower Vessel Head of a Nuclear Reactor

This paper describes progress on a consistent approach for multi-phase flow modelling with phase-change. Although, the de- veloped methods are general purpose the applications presented here cover the LIVE experiments involving core melt phenomena at the lower vessel head of a nuclear reactor. These include corium pool formation, coolability and solidification. A new method for solving the compositional multi-phase flow equations to calculate material indicator fields is adopted. An interface-capturing scheme based on high-order accurate compressive advection methods including a Petrov-Galerkin approach is employed to main- tain sharpness of the interfaces between materials. A novel control volume-finite element mixed formulation based on the P1DG-P2 (linear discontinuous in velocity and quadratic continuous in pressure) element pair which can uniquely ensure that key balances are maintained in the equations is developed to discretise the governing equations in space. Anisotropic mesh adaptivity is used to focus the numerical resolution around the interfaces and other areas of important dynamics.

Analysis of accidental loss of pool coolant due to leakage in a PWR SFP

A large loss of pool coolant/water accident may be caused by extreme accidents such as the pool wall or bottom floor punctures due to a large aircraft strike. The safety of SFP under this circumstance is very important. Large amounts of radioactive materials would be easily released into the environment if a severe accident happened in the SFP, because the spent fuel pool (SFP) in a PWR nuclear power station (NPS) is often located in the fuel handing building outside the reactor containment. To gain insight into the loss of pool coolant accident progression for a pressurized water reactor (PWR) SFP, a computational model was established by using the Modular Accident Analysis Program (MAAP5). Important factors such as Zr oxidation by air, air natural circulation and thermal radiation were considered for partial and complete drainage accidents without mitigation measures. The calculation indicated that even if the residual water level was in the active fuel region, there was a chance to effectively remove the decay heat through axial heat conduction (if the pool cooling system failed) or steam cooling (if the pool cooling system was working). For sensitivity study, the effects of emergency ventilation and water injection on the accident progression were analyzed. The analysis showed that for the current configuration of high-density storage racks, it was difficult to cool the spent fuels by air natural circulation. Enlarging the space between the adjacent assemblies was a way of increasing air natural circulation flow rate and maintaining the coolability of SFP. Water injection to the bottom of the SFP helped to recover water inventory, quenching the high temperature assemblies to prevent melting, and even the molten corium pool to avoid molten-corium-concrete-interaction (MCCI).

Detailed evaluation of melt pool configuration in the lower plenum of the APR1400 reactor vessel during severe accidents

For a detailed evaluation of the IVR (In-Vessel corium Retention) through the ERVC (External Reactor Vessel Cooling) during a severe accident, the melt pool configuration should be accurately determined in the lower plenum of the reactor vessel. It affects the thermal load to the vessel wall and plays a key role in determining the integrity of the reactor vessel under the IVR-ERVC. SCDAP/RELAP5 and GEMINI analyses have been performed to determine the final corium condition and examine the final melt pool composition at a reactor vessel failure during a severe accident in an APR (Advanced Power Reactor) 1400, respectively. As the representative accident scenarios, five dominant sequences of the TLFW (Total Loss of Feed Water), the SBO (Station BlackOut), the SBLOCA (Small Break Loss of Coolant Accident) without SI (Safety Injection), the MBLOCA without SI, and the LBLOCA without SI were selected from the level I PSA (Probabilistic Safe Assessment) results. A density evaluation graph was developed for the precise examination of the melt pool layer inversion. The final melt pool configurations at the reactor vessel failure have been determined for five dominant accident scenarios of the APR1400 using the GEMINI results and the density evaluation graph. The thermodynamic analysis results in three sequences of the APR1400 accident address the possibility of a melt pool layer inversion in the lower plenum of the reactor vessel. The layer inversion led to corium pool stratification with a heavy metallic layer below the oxidic pool, which leads to a three-layer formation.

Oxidation effect on steel corrosion and thermal loads during corium melt in-vessel retention

During a severe accident with core meltdown, the in-vessel molten core retention is challenged by the vessel steel ablation due to thermal and physicochemical interaction of melt with steel. In accidents with oxidizing atmosphere above the melt surface, a low melting point UO2+x–ZrO2–FeOy corium pool can form. In this case ablation of the RPV steel interacting with the molten corium is a corrosion process.Experiments carried out within the International Scientific and Technology Center's (ISTC) METCOR Project have shown that the corrosion rate can vary and depends on both surface temperature of the RPV steel and oxygen potential of the melt. If the oxygen potential is low, the corrosion rate is controlled by the solid phase diffusion of Fe ions in the corrosion layer. At high oxygen potential and steel surface layer temperature of 1050°C and higher, the corrosion rate intensifies because of corrosion layer liquefaction and liquid phase diffusion of Fe ions.The paper analyzes conditions under which corrosion intensification occurs and can impact on in-vessel melt retention (IVR).

An experimental study on layer inversion in the corium pool during a severe accident

COSMOS (COrium configuration of the molten State in the MOst Severe accidents) tests using prototypic materials have been performed to investigate the layer inversion of the heavy metallic material in the corium pool. An induction heating method using a cold crucible was implemented for melting of the UO2–ZrO2–Zr–Fe mixture. Before the main test, three preliminary tests have been performed to enhance the experimental techniques using the real core material. One main test of the COSMOS has been performed to evaluate the corium pool configuration in the lower plenum of the reactor vessel for the TLFW (Total Loss of Feed Water) sequence of the APR (Advanced Power Reactor) 1400 under the IVR-ERVC (In-Vessel corium Retention through External Reactor Vessel Cooling). A post-test examination of cutting, EPMA, and XRD for the solidified corium pool ingot in the main test has been performed to investigate the layer inversion. From the three preliminary tests, melting techniques of the real core material using a cold crucible were successfully developed. The metallurgical inspection results on the chemical information coincide with the visual observation on the centerline cut ingot in that the upper part is metal and the lower lump is an oxidic mixture with some metal clods in the main test. This means the possibility of layer inversion of the heavy metallic material in the corium pool.

Research on the possibility of material mixing and interaction in the lower head based on specific corium relocation process

MASCA experiment indicated the possibility of oxidic and metallic material interaction and the following analysis produced a possible 3-layer configuration in the lower head. However, 3-layer corium pool is valid only under the circumstance that materials are well mixed. This assumption has not been studied thoroughly in IVR analysis. In this paper, a CFD code is used to predict the possibility of material mixing based on the MAAP4 results and conservative assumptions. The analysis indicates that the validated LES model and a grid of y+<10 can well predict thermal hydraulics in the oxide and metal phase, thus producing boundary conditions for the crust integrity analysis. The stress in the crust is calculated by ANSYS code to determine the integrity of crust by comparison with MACE experiments. The results show that the crust is likely to maintain its integrity and keep the oxidic and metallic materials separated, thus restricting the material interaction. This study may support a 2-layer corium pool configuration in the lower head.

THERMAL AND STRUCTURAL ANALYSIS OF CALANDRIA VESSEL OF A PHWR DURING A SEVERE ACCIDENT

In a postulated severe core damage accident in a PHWR, multiple failures of core cooling systems may lead to the collapse of pressure tubes and calandria tubes, which may ultimately relocate inside the calandria vessel forming a terminal debris bed. The debris bed, which may reach high temperatures due to the decay heat, is cooled by the moderator in the calandria. With time, the moderator is evaporated and after some time, a hot dry debris bed is formed. The debris bed transfers heat to the calandria vault water which acts as the ultimate heat sink. However, the questions remain: how long would the vault water be an ultimate heat sink, and what would be the failure mode of the calandria vessel if the heat sink capability of the reactor vault water is lost? In the present study, a numerical analysis is performed to evaluate the thermal loads and the stresses in the calandria vessel following the above accident scenario. The heat transfer from the molten corium pool to the surrounding is assumed to be by a combination of radiation, conduction, and convection from the calandria vessel wall to the vault water. From the temperature distribution in the vessel wall, the transient thermal loads have been evaluated. The strain rate and the vessel failure have been evaluated for the above scenario.

Evaluation of in-vessel corium retention through external reactor vessel cooling for small integral reactor

A criterion to prevent reactor vessel failure using in-vessel corium retention through the external reactor vessel cooling (IVR-ERVC) of a small integral reactor has been evaluated during a severe accident. A thermal load analysis from the corium pool to the outer reactor vessel wall in the lower plenum of the reactor vessel was performed to determine the heat flux distribution. The critical heat flux (CHF) on the outer reactor vessel wall was determined to fix the maximum heat removal rate through the external coolant between the outer reactor vessel and the insulation of the reactor vessel. Finally, the thermal margin was evaluated through a comparison of the thermal load with the maximum heat removal rate of the CHF on the outer reactor vessel wall. The maximum heat flux from the corium pool to the outer reactor vessel is estimated as approximately 0.25MW/m2 in the metallic layer, owing to the focusing effect. The CHF of the outer reactor vessel is approximately 1.0–1.1MW/m2 because of a two-phase natural circulation mass flow rate. Since the thermal margin for the IVR-ERVC is sufficient, the reactor vessel integrity is maintained during a severe accident in the small integral reactor.

Oxidation kinetics of corium pool

Experimental, theoretical and numerical studies of oxidation kinetics of an open surface corium pool have been reported. The experiments have been carried out within OECD MASCA program and ISTC METCOR, METCOR-P and EVAN projects. It has been shown that the melt oxidation is controlled by an oxidant supply to the melt free surface from the atmosphere, not by the reducer supply from the melt. The project experiments have not detected any input of the zirconium oxidation kinetics into the process chemistry. The completed analysis puts forward a simple analytical model, which gives an explanation of the main features of melt oxidation process. The numerical modeling results are in good agreement with experimental data and theoretical considerations.

LIVE-L4 and LIVE-L5L Experiments on Melt Pool and Crust Behavior in Lower Head of Reactor Pressure Vessel

The LIVE-L4 and LIVE-L5L experiments investigated the thermal-hydraulic behavior of the corium pool in the reactor pressure vessel lower head with the three-dimensional test vessel LIVE. The simulant material is a noneutectic binary mixture of 20% NaNO3-80% KNO3. Transient and steady-state parameters such as melt temperature and heat flux distribution through the vessel wall as well as crust formation characteristics were obtained. The two tests demonstrated that transient events like repeated melt relocation and change of decay power density facilitate crust deformation and change of crust thickness. Massive crust formation in a noneutectic melt pool leads to a change of melt pool composition and a decrease of melt-crust interface temperature. The melt temperature and heat flux at the same pool height and same power density can be roughly compared independent of heating history and initial melt pouring pattern. The dimensionless melt temperature as well as the dimensionless heat flux through the wall during the steady state are independent of power density if the pools have the same height. But, they are dependent on the pool height. For a low pool, the gradients with height of both melt temperature and heat flux through the vessel are larger than those for a high pool.

Numerical analysis of debris melting phenomena during late phase CANDU 6 severe accident

The objective of this paper was to study the phase change of the debris formed on the CANDU 6 calandria bottom in a postulated accident sequence. The molten pool and crust formation were studied employing the Ansys Fluent code. Two models using Large Eddy Simulation and Phase Change Effective Convectivity Model (PECM) predict the conjugate, radiative and convective heat transfer inside and from the corium pool. LES require a very fine grid to capture the crust formation and the free convection flow. This aspect (fine mesh requirement) correlated with the long transient has imposed the use of a slice from the 3D calandria geometry in order not to exceed the computing resources. From the safety point of view it is very important to maintain a heat flux through the wall below the critical heat flux assuring the integrity of the calandria vessel, thus the paper results include temperature profiles and heat fluxes through calandria wall. The results predicted by the PECM approach are comparable within 10% with those obtained by the computational fluid dynamics method, proving that the PECM can be used to perform further sensitivity studies.

CORIUM BEHAVIOR IN THE LOWER PLENUM OF THE REACTOR VESSEL UNDER IVR-ERVC CONDITION: TECHNICAL ISSUES

Corium behavior in the lower plenum of the reactor vessel during a severe accident is very important, as this affects a failure mechanism of the lower head vessel and a thermal load to the outer reactor vessel under the IVR-ERVC (In-Vessel corium Retention through External Reactor Vessel Cooling) condition. This paper discusses the state of the art and technical issues on corium behavior in the lower plenum, such as initial corium pool formation characteristics and its transient behavior, natural convection heat transfer in various geometries, natural convection heat transfer with a phase change of melting and solidification, and corium interaction with a lower head vessel including penetrations of the ICI (In-Core Instrumentation) nozzle are discussed. It is recommended that more detailed analysis and experiments are necessary to solve the uncertainties of corium behavior in the lower plenum of the reactor vessel.

CONTRIBUTIONS OF THE VULCANO EXPERIMENTAL PROGRAMME TO THE UNDERSTANDING OF MCCI PHENOMENA

Molten Core Concrete Interaction (MCCI) is a complex process characterized by concrete ablation and volatile generation; Thermal and solutal convection in a bubble-agitated melt; Physico-chemical evolution of the corium pool with a wide solidification range (of the order of 1000 K). Twelve experiments have been carried out in the VULCANO facility with prototypic corium and sustained heating. The dry oxidic corium tests have contributed to show that silica-rich concrete experience an anisotropic ablation. This unexpected ablation pattern is quite reproducible and can be recalculated, provided an empirical anisotropy factor is assumed. Dry tests with oxide and metal liquid phases have also yielded unexpected results: a larger than expected steel oxidation and unexpected topology of the metallic phase (at the bottom of the cavity and also on the vertical concrete walls). Finally, VULCANO has proved its interest for the study of mitigation solutions such as the COMET bottom flooding core catcher.

Numerical analysis of core catcher efficiency for VVER-1200

A brief description of the HEFEST-ULR code and numerical results for corium behavior in a core catcher for the VVER-1200 (NPP-2006 project) are presented. Physical, chemical, and thermal processes during corium pool formation and cooling in the core catcher are examined for the case of severe accidents in NPP-2006 with the VVER-1200. Basic corium parameters, including its composition, structure, temperature evolution, density, and aggregative state, are determined. The minimum margin for critical heat flux at the external surface of the core catcher is estimated.

A Study on Natural Circulation Flow under Reactor Cavity Flooding Condition in Advanced PWR

Cavity flooding is an important severe accident management measure for in-vessel retention of a degraded core by external reactor vessel cooling in advanced PWRs. A code simulation study on the natural circulation flow in the gap between the reactor vessel wall and insulation material under cavity flooding condition is performed by using a detailed mechanistic thermal-hydraulic code package RELAP 5. By simulating of an experiment carried out for studying the natural circulation flow for APR 1400 shows that the code is applicable for analyzing the circulation flow under this condition. Molten corium pool in vessel lower plenum imposes highest heat load on the vessel lower head wall in large-break loss of coolant accident. The analysis results show that heat removal capacity of the natural circulation flow in AP1000 is sufficient to prevent thermal failure of the reactor vessel under this bounding heat load. By sensitivity analysis, some insights are provided for increasing the thermal margin between the heat source in the corium pool of the reactor lower plenum and the heat transfer from the lower vessel wall to the coolant in the gap by improvement of the vessel/insulation configurations.

Interface temperature between solid and liquid corium in severe accident situations: A comprehensive study of characteristic time delay needed for reaching liquidus temperature

Tliquidus was proposed as the interface temperature (Seiler and Froment, 2000), for various severe accident situations for thermalhydraulic steady state. This proposal was made on the basis of the analysis of solidification front stability in thermalhydraulic steady state for volumetrically heated corium pools and was extended to reactor transients with slow solidification rates that are controlled by the longterm decrease in residual power. The conclusions were corroborated by prototypic corium and variable solidification rates obtained by experimental approaches (Dauvois et al., 2000; Journeau et al., 2003) for corium containing small amounts of silica or none at all. When the concentration in silica increases (approximately above 10 wt%), it was concluded from the experiments that a plane-front situation could not be obtained. The present work offers a theoretical approach to the maximum time delay that is necessary for mass transfer and full phase-segregation in volumetrically heated liquid pools bounded by a crust. It is concluded that full segregation is obtained for in-vessel situations within time delays that are shorter or of the same order of magnitude as the characteristic time for the corium pool to form and evolve to a quasi-steady-state situation. The characteristic time delay for mass transfer associated with simulant material experiments is also determined. Phase segregation can also be obtained for corium–concrete interaction, provided that the silica content is less than approximately 10 wt%. However in the latter case, more complex phenomena occur at the interface due to the interaction with sparging gas (such as porous medium formation) which requires a different model approach.

Transient Conduction Heat Transfer Modeling in Concrete for the Simulation of Long-Term Phase of Molten Core-Concrete Interaction

Heat transfer between corium pool and concrete directly governs the ablation velocity of concrete in the case of molten core-concrete interaction (MCCI) and, consequently, the time delay when the reactor cavity may fail. Numerical tools dealing with MCCI generally consider that the ablation velocity of concrete is higher than the “velocity” of heat transfer inside the concrete so that conduction heat transfer in the basemat is not taken into account. With such modeling, concrete ablation goes on until the heat flux between the corium pool and the concrete is zero. This assumption proved to be satisfactory for high heat flux because of the low thermal diffusivity of concrete. Nevertheless, it can be discussed in cases where the heat flux between the corium and the concrete is “low” that is in the long-term phase of MCCI or in cases with a strong imbalance in the power splitting at the corium pool boundaries. In such situations, the heat transfer by conduction in the concrete is no longer negligible and can lead to the end of the concrete ablation. Heat conduction in the concrete could be taken into account by solving multi-dimensional transient heat transfer equations in the concrete. A spatial meshing of the basemat is then necessary, but such an approach is time-consuming. That is why a simplified one-dimensional transient approach has been chosen and implemented in the TOLBIAC-ICB code. The main purpose of this paper is to present this approach. The validation has been performed by comparing the results of this method with experimental data obtained from studying the thermal response of polymethylmetacrylate and concrete to a heat flux. Results of the model are also compared to the solutions obtained by the numerical resolution of the discretized heat transfer equation on a fine mesh. Finally, an application to the reactor case is proposed.