Simulation of Core Melt Pool Formation in a Reactor Pressure Vessel Lower Head Using an Effective Convectivity Model (original) (raw)
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Dynamics of heat transfer in the melt pool at nuclear severe accident conditions
Prediction of thermal loads on nuclear reactor vessel lower plenum after core melting and relocation during a severe accident requires knowledge about the core melt behavior, especially the circulation pattern. To analyze the heat transfer dynamics on the lower plenum walls, two-dimensional numerical simulations of a fluid flow with internal heat generation were performed for Rayleigh numbers 10^6, 10^7, 10^8, 10^9, 10^11 and 10^13 at Prandtl number 0.8. For subgrid motion modeling, a Large-Eddy Simulation Smagorinsky model was implemented. The minimum, time-average and maximum Nusselt numbers on the boundaries were calculated. The dynamics of fluid structures were analyzed to reveal the instability mechanisms and transition to turbulence. Results disclose Rayleigh-Taylor instabilities as a dominant mechanism for turbulence appearance, which occurs when the Rayleigh number is increased over 10^8. The structure dependence of fluid motion at high Rayleigh numbers makes the time-average of heat transfer hard to assess. The time-average values should be supplemented with probability distributions of related variables.
Dynamic behavior of the melt pool at severe accident conditions
Prediction of thermal loads on lower plenum walls after core melting and relocation during severe accident conditions requires knowledge about the core melt behavior, especially the circulation pattern. To analyze the heat transfer dynamics on the lower plenum walls, two-dimensional numerical simulations of a fluid flow with internal heat generation were performed for Rayleigh numbers 10^6, 10^7, 10^8, 10^9, 10^11 and 10^13 at Prandtl number 0.8. For subgrid motion modeling, the Large Eddy Simulation (LES) Smagorinsky model was implemented. Time and boundary-averaged Nusselt numbers were calculated. Results show that differences between minimum, average and maximum Nusselt number increase in exponential manner when the Rayleigh number is increased beyond 108. Probability densities of Nusselt number were also calculated to realistically assess unsteady thermal loads. The calculated probability density functions indicate that time-average Nusselt numbers usually do not coincide with most probable values. The study also discloses the appearance of multiple Nusselt number probability peaks.
A possible severe accident scenario is a general meltdown and relocation of the reactor core during which molten core material accumulates in the lower plenum of the reactor vessel. The decay heat generated in a radioactive material would have to be removed through the walls of the lower plenum in order to ensure the integrity of the reactor pressure vessel. Numerical simulations of turbulent natural convection in a geometry representing the lower plenum cavity of a reactor pressure vessel were conducted. A two-dimensional numerical code based on a finite-volume method was developed to simulate turbulent natural convection in a fluid with internal heat generation using large-eddy simulation. Simulations were performed at Rayleigh numbers 1e10 and 2e11 and Prandtl numbers 1.2, 7 and 8, which corresponds to conditions in the numerical investigations made by Nourgaliev et al. (1997) and in the experimental work done by Asfia and Dhir (1996). The results are shown to be in satisfactory agreement.
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...
Nuclear Engineering and Design, 1996
The objective of this paper is to study the heat and mass transfer process~ rdated ~o core melt discharge from a reactor vessel in a light water reactor severe accident. The phenomenology modell~c~ includes the convection in, and heat transfer from. the melt pool in contact with ~he vessel lower head wall, the fluid dynamics and heat transfer of the melt flow in the growing discharge hole and multi-dimensional heat conduction in the ablating lower head wall. A research programme is underway at the Royal Institute of Technology (Kungliga Tekniska H6gskolan, KTH) to (t) identify the dominant heat and mass transfer processes determining the characteristics of the lower head ablation process: (2) develop and vaEda~c c~dent anaiytieat/eomputational models for these processes; (3) apply models to assess the character of the melt discharge process in a reactor-scale situation; (4) determine the sensitivity of the melt discharge to structural differences and variations in the in-vesscl melt progression scenarios. The paper also presents a comparison with recent results of vessel hole ablation experiments conducted at KTH with a meR simulant.
LIVE Experiments on Melt Behavior in the Reactor Pressure Vessel Lower Head
Heat Transfer Engineering, 2013
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.
Modeling of natural convection phenomena in nuclear reactor core melt
A failure of reactor core cooling and major safety systems may cause melting of nuclear fuel and reactor vessel equipment. In the reactor vessel, the melt flows down into the lower plenum, where it is accumulated. In the past, the common opinion was that the melt would break through the reactor vessel and start to desintegrate the reactor concrete base. However, recent investigations revealed that the core melt can be safely retained in the reactor vessel lower plenum if it is properly cooled. The processes in the reactor core melt in the lower plenum are specific and not yet fully understood. As revealed by a comprehensive overview of the lower plenum cooling problem, natural convection is the most important phenomenon that controls heat transfer from the melt. In the case of natural convection, fluid motion is caused by volumetric forces and density gradients. If these are strong enough, thermal instabilities may result in hydrodynamic instabilities. It was discovered that transition from laminar to turbulent flow occurs at the value of Rayleigh number Ra=5e5 in the case of Rayleigh-Bénard convection and at Ra=1e6 in the case of fluid with internal heat generation. The main problem of turbulent phenomena modeling is the size of turbulent fluid flow structures, which are in general too small to be described accurately using a discrete numerical mesh. The base of the Smagorinsky model is the assumption that the smallest flow structures, which are separated and modeled as a subgrid term, are isotropic and homogeneous. Therefore, viscous dissipation is equal to the production of turbulent kinetic energy. As the Smagorinsky model is too dissipative in the vicinity of the walls, turbulent viscosity wall functions have to be implemented. Natural convection in the melt of nuclear reactor core was modelled as natural convection in a fluid with internal heat generation in a rectangular cavity. The value of Rayleigh number was Ra=1e10 and the value of Prandtl number was Pr=1.2. Numerical simulations were restricted to two-dimensional space, due to computer hardware limitations. The finite volume method was used for spatial discretisation and a combination of Adam-Bashford method and projection scheme was used for time integration. As the calculation of heat transfer in the form of dimensionless Nusselt number revealed, the most severe thermal loads occur on the side walls in the vicinity of the cavity upper boundary. Calculated values of heat transfer can be safely extrapolated to higher values of Rayleigh number.
2015
The development of analysis models for the Swiss reactors is a key objective of the STARS project at the Paul Scherrer Institut (PSI). Within this context there is a need for the development of computational fluid dynamics (CFD) models of the Swiss reactors in support of future high fidelity investigations of steadystate and transient scenarios. This article presents initial results for the CFD analysis of a Siemens KWU PWR with a focus on the flow behavior and heat transfer in the gap between the core shroud and core barrel. Temperatures and densities in this region of the reactor are important for accurate estimations of fast neutron fluence and activation in the steel structures of the core shroud, core barrel and reactor pressure vessel. The flow behavior in this region may also be relevant in the understanding of ex-core detector responses. The flow conditions in the core bypass region were found to be in the transition-toturbulence regime, with vortex shedding taking place dow...
Simulation of Core Melt Progression Using a Simulant Metal Alloy
2002
Numerous experiments were conducted to address when and how the core can lose its original geometry, what geometries are formed, and in what processes the core materials are transported to the lower plenum of the reactor pressure vessel during a severe accident. Core degradation progresses along the line of clad ballooning, clad oxidation, material interaction, metallic blockage, molten pool formation, melt progression, and relocation to the lower head. Relocation into the lower plenum may occur from the lateral periphery or from the bottom of the core depending upon the thermal and physical states of the pool. Determining the quantities and rate of molten material transfer to the lower head is important since significant amounts of molten material relocated to the lower head can threaten the vessel integrity by steam explosion and thermal and mechanical attack of the melt. In this paper the focus is placed on the melt flow regime on a cylindrical fuel rod utilizing the LAMDA (Lumped Analysis of Melting in Degrading Assemblies) facility at the Seoul National University. The downward relocation of the molten material is a combination of the external film flow and the internal pipe flow. The heater rods are 0.8 m long and are coated by a low-temperature melting metal alloy. The electrical internal heating method is employed during the test. External heating is adopted to simulate the exothermic Zircaloy-steam reaction. Tests are conducted in several quasi-steady-state conditions. Given the variable boundary conditions including the heat flux and the water level, observation is made for the melting location, progression, and the mass of molten material. Finally, the core melt progression model is developed from the visual inspection and quantitative analysis of the experimental data. As the core material relocates downwards a blockage may be formed and grow both radially and axially. The velocity of the melt can be calculated from a force balance between the gravity and frictional losses at the melt-rod interface. When the heater rod is uncovered completely, the melt progression is initiated at the mid-point, which is the hot spot in the rod. However, the melting location is elevated as the water level rises because of the downward heat transfer. Considering the melt flow as a film, the steady-state film thickness on the cylindrical heater rod and the average velocity are computed. The steady-state film flow rate is determined in terms of the density, film thickness, and film velocity.
CFD analysis of PWR core top and reactor vessel upper plenum internal subdomain models
Nuclear Engineering and Design, 2011
One aspect of the Westinghouse AP1000 TM 1 reactor design is the reduction in the number of major components and simplification in manufacturing. One design change relative to current Westinghouse reactors of similar size is that AP1000 reactor vessel has two nozzles/hot legs instead of three. With regard to fuel performance, this design difference creates a different flow field in the reactor vessel upper plenum. The flow exiting from the core and entering the upper plenum must turn toward one of the two outlet nozzles and flow laterally around numerous control rod guide tubes and support columns. Also, below the upper plenum are the upper core plate and the top core region of the 157 fuel assemblies and 69 guidetube assemblies.