Direct Numerical Simulation of Fluid Flow in a 5x5 Square Rod Bundle Using Nek5000

Rod bundle flows are commonplace in nuclear engineering, and are present in light water reactors (LWRs) as well as other more advanced concepts. Inhomogeneities in the bundle cross section can lead to complex flow phenomena, including varying local conditions of turbulence. Despite the decades of numerical and experimental investigations regarding this topic, and the importance of elucidating the physics of the flow field, to date there are few publicly available direct numerical simulations (DNS) of the flow in multiple-pin rod bundles. Thus a multiple-pin DNS study can provide significant value toward reaching a deeper understanding of the flow physics, as well as a reference simulation for development of various reduced-resolution analysis techniques. To this end, DNS of the flow in a square 5 × 5 rod bundle at Reynolds number of 19,000 has been performed using the highly-parallel spectral element code Nek5000. The geometrical dimensions were representative of typical LWR fuel designs. The DNS was designed using microscales estimated from an advanced Reynolds-Averaged Navier-Stokes (RANS) model. Characteristics of the velocity field, Reynolds stresses, and anisotropy are presented in detail for various regions of the bundle. The turbulent kinetic energy budget is also presented and discussed.

Experimental Study on Natural Convection Heat Transfer Outside Tube Bundle in Space Under Low Temperature Difference

The Passive Residual Heat Removal System (PRHRS) is very important for the safety of the heating reactor after shutdown. PRHRS is a natural circulation system driven by density difference, therefore, the heat transfer performance of the Passive Residual Heat Removal Heat Exchanger (PRHR HX) has a great impact to the heat transfer efficiency of PRHRS. However, the most research object of PRHR HX is the C-shape heat exchanger at present, which located in In-containment Refueling Water Storage Tank (IRWST). This heat exchanger is mainly used for the PRHRS of nuclear power plants. In the swimming pool-type low-temperature heating reactor (SPLTHR), the PRHR HX is placed in the reactor pool, which the pressure and temperature of the reactor pool are relatively low, and the outside heat transfer mode of tube bundle is mainly natural convection heat transfer. In this study, a miniaturized single-phase pool water cooling system was built to investigate the natural convective heat transfer coefficient of the heat exchanger under the large space and low temperature conditions. The experimental data had been compared with several correlations. The results show that the predicted value of Yang correlation is the closest to the experimental data, which the maximum deviation is about 11%.

A Comparative Study of Constrained and Unconstrained Melting Inside a Sphere

Present study is focused on the computational analysis of melting of PCM inside the spherical capsule. Both unconstrained and constrained melting is analyzed for the constant PCM volume and similar initial and boundary conditions. RT27 is chosen as the PCM for this study. Air is considered at the top of PCM inside the spherical capsule. Results are validated with the existing experimental and computational results and found to be in good agreement. Results obtained from present study are compared for the melting fraction, pattern and time. Composite diagrams are presented for the streamline and temperature contours.

Thermal Safety Margin Calculation of the MP-2 Experiment in the Advanced Test Reactor

The Mini-Plate 2 (MP-2) irradiation test is a fueled experiment designed for irradiation in multiple test locations in the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL). The experiment is a drop-in test where small aluminum-clad fuel plate samples (mini plates) are cooled directly by the ATR Primary Coolant System (PCS) water. The MP-2 fuel plate experiment will be irradiated in several different irradiation locations of the ATR. This fueled experiment contains aluminum-clad fuel mini plates consisting of monolithic U-Mo. Four different types of fuel plates with fuel meat thickness and cladding are part of the MP-2 test. A thermal analysis has been performed on the MP-2 experiment. A method for calculating Departure from Nucleate Boiling Ratio (DNBR) and Flow Instability Ratio (FIR) during a reactivity transient using the commercial finite element and heat transfer code ABAQUS is discussed. At the start of an ATR cycle the heat generation rate of the fueled experiment is high and the heat rate multiplier from the outer shim control cylinders is low, while the reverse is true at the end of the ATR cycle. Thermal analyses at 10-day increments during the cycle calculate the DNBR and FIR during a reactivity transient. This technique calculates DNBR for the fuel plate surfaces and FIR for all water components for each finite element surface and node at various times during the ATR cycle. Heat rates vary with time during the transient calculations that are provided by a detailed physics analysis. Oxide growth on the fuel plates is also incorporated. Results from the transient calculations are displayed with the ABAQUS post processor. By calculating these parameters at each location in the finite element model, conservatism is replaced with accuracy. This allows for a greater margin for the thermal hydraulic safety parameters.

Simulations of Experiments on Isothermal Containment Atmosphere Mixing Caused by Vertical Injection

An experiment performed in SPARC experimental facility, was simulated with the computational fluid dynamics code OpenFOAM. The experiment took place in two phases. In the first phase, a helium-air layer was generated, which was then eroded with a vertical air jet injected in the vessel axis during the second experimental phase. A three-dimensional and a quasi-two-dimensional numerical models of a cylindrical vessel were developed and mesh convergence studies were performed. The mixing process was simulated as a single-phase flow with common momentum equation. Included gas species mass fractions are considered as passive scalars and are calculated using the transport equation. However, term describing the molecular diffusion cannot be neglected in our case and had to be added to diffusion equation implemented in default OpenFOAM solver. The k-ε turbulence model with additional buoyancy term implemented in OpenFOAM was used for turbulence modelling. Despite improvement of the physical model and following the Best Practice Guidelines, the results obtained with OpenFOAM CFD code still, at some locations, differ substantially from the experimental results. A modified definition of low Reynolds number eddy viscosity correction function is proposed, which significantly improves the agreement between measurements and calculation results.

The Dependence of Pool Scrubbing Decontamination Factor on Particle Number Density: Modeling Based on Bubble Mass and Energy Balance

The dependence of pool scrubbing performance on particle number density is studied through numerical simulation of experimental results. The DF values obtained from the authors’ experiments (Sun et al., Sci. Technol. Nucl. Inst., Article ID 1743982, 2019) indicate a sharp decrease with an increase in the inlet particle number density beyond 1011/m3. The mechanisms underlying such dependence is yet to be studied. In this paper, a simple model is developed to study the factors affecting the experimentally observed dependence of DF. The test results suggest that the condensational growth of particles plays an essential role in the inertial capture. The vapor condensation on the particles has an effect to deplete the vapor supersaturation in the bubble by both lowering the vapor concentration and raising the temperature. This effect will become important at high particle number densities. The bubble mass and energy balance is calculated to derive the particle growth and the inertial DF as a function of the bubble rise distance through the pool water. The balance is assumed to be quasi-steady, and the vapor concentration and the temperature to be uniform in the bubble. It is shown that the model reproduces the tendency observed in the experimental DF. The model predicts that the degree of supersaturation is affected when particle concentration exceeds 1011/m3, curbing the condensational growth of particles, and thereby retarding the inertial capture.

Dynamic PRA Based on System Codes Coupling for Passive Safety System in Integral Pressurized Water Reactor

An integral Pressurized Water Reactor (iPWR) type SMR was studied featuring Passive Safety Systems (PSSs). Different from active systems, PSSs are easily influenced by system parameters referred to as phenomenological factors such as heat loss, flow friction, oxidation, non-condensable gases and void fraction due to the low driving force of natural circulation. The system parameters also contribute to the uncertainty and dependency of PSS leading to the system unreliability. Thus, efforts are made to improve the reliability of PSS. A classical Probabilistic Risk Assessment (PRA) model describing active systems does not consider time evolution nor event ordering for PSS that dynamic PRA can accommodate. Here we developed and realized coupling between LabVIEW and CAFTA. Isolation Condenser System (ICS) is taken as the benchmark system due to the simple design in single phase without phase change phenomena in order to mainly remove decay heat and secondarily depressurize the reactor pressure vessel (RPV). A classical PRA model of ICS using CAFTA is coupled with real-time simulation of primary loop and ICS in LabVIEW, leading to a dynamic simulation result. The difference in failure probability using dynamic versus classical PRA revealed that for one there are more component demands with different event ordering, such that improved PSS reliability in the iPWR-type SMR designs is possible.

Thermal-Hydraulic Design Support and Safety Analyses of SEALER UK Demo

SEALER (SwEdish Advanced Lead Reactor) is a passively safe lead-cooled reactor designed for commercial power production, under design by the LeadCold company. The reactor is modular in design, allowing for factory production and reduction in investment risk compared with new-build of large Light Water Reactors. Furthermore, its core is designed such that it can generate power for up to 25 years without the need of on-site fuel-cycle operations. The SEALER UK model has specifically been designed to produce base-load power on the UK grid. In the design and safety evaluation process, NRG is currently providing support to LeadCold Reactors with respect to thermal-hydraulic safety analyses utilizing Computational Fluid Dynamics (CFD) competences. The current paper gives a comprehensive description of a 3D CFD model created of SEALER UK Demo, which is a scaled-down demonstrator of SEALER UK. The geometry of the CFD model of SEALER UK Demo as well as the modelling approach and numerical settings are presented here. Assumptions were made in order to make it computationally feasible to perform simulations. These are discussed as well. Subsequently, the 3D CFD model is used to perform steady-state analyses of SEALER UK Demo operating under nominal conditions. Main parameters such as mass flow rates, temperatures and core pressure drops coming from the model match the design values well, with differences being at most a couple percent. Also, it is found that the margin to lead freezing with the current design parameters is more than 50K.

Applicability of Sub-Modelling Technique for Dynamic Analysis of Concrete Structures With Attached Equipment Under Missile Impact

This paper describes the work conducted by the Canadian Nuclear Safety Commission (CNSC) related to the numerical simulations of reinforced concrete (RC) structures under deformable missile impact. The current paper is a continuation of the work conducted in the frame of the OECD/NEA* IRIS (Improving Robustness Assessment Methodologies for Structures Impacted by Missiles) Phase 3 benchmark project. The concrete mock-up with two simple structures attached, one welded and another bolted, was built and tested at the VTT Technical Research Centre in Espoo, Finland. This mock-up was impacted by three subsequent missiles with varying velocities in order to obtain the damage accumulation. To examine vibration transmission through the mock-up, the simple structures modelling equipment were attached to the rear wall of the structure, while the missile impact was at the centre of the front wall. The parameters of the missiles and the RC structure were selected to ensure a flexible behaviour of the RC target in the impact area with only moderate damages, specifically cracking and permanent deformation without perforation. The non-linear dynamic behaviour of the reinforced concrete slabs under missile impact was analyzed using the commercial FE code LS-DYNA. A hybrid FE model using both 3-D solid and 2-D shell FE models was developed for the target discretization. Since the ultimate objective of this work is to model the entire structure over long time periods, a simplified combined shell-solid model with distributed (smeared) reinforcement was selected and validated. This model employs solid FE around an impact area and shell FE for the rest of the mock-up. Detailed modelling of a large RC structure with all equipment attached leads to a very large finite element (FE) model. Therefore, two-level FE modelling using sub-modelling approach was employed: first, analyze the vibrations of a reinforced concrete structure with simplified equipment modelling, and second, analyze in detail the equipment connected to it. This approach assumes uncoupled dynamic behaviour of the structure and the equipment. While the sub-modelling technique is commonly used in static analysis, a special sensitivity analysis was conducted to prove the applicability of sub-modelling for impact analysis. Finally, the effect of structural damping was examined and the best possible damping was selected. The selected damping values and sub-models resulted in relatively good agreement with the test results for both global (RC mock-up) and local (equipment) behaviour.

Neutronics and Thermal Hydraulics Coupled 3-D Calculation of Pressure-Tube Deformation

Among the various parts in a pressurized heavy-water reactor (PHWR), pressure tubes are of tremendous importance. This is because they withstand extreme both pressure and temperature differences that exist between the Primary Heat Transport System (PHTS) and the moderator. The pressure tubes also contribute to prevention of fission product release from the PHTS to the PHWR plant (together with end fittings and nearby parts including plugs). When a PHWR is given a 1% derating, half is due to the aging of the pressure tubes. The main concern with pressure tubes is decrease of the safety margin. Most of the reduction comes from the effects caused by radial expansion and axial sagging, which are belong to four major phenomena including the thinning and the elongation. More specifically, the fuel-pin temperature distribution changes for the worse if deformation of the pressure tube occurs. Because there is extreme irradiation inside the core, the tube content is exposed to high temperature and high pressure. Thus, the shape of the pressure tube is deformed as times goes on. In this paper, using modeling of a deformed pressure tube in three-dimensional space, the effects on the fuel, coolant temperature, and coolant density, were studied quantitatively. This included a neutronics effect explored using coupled neutronics and thermal hydraulics (T/H) calculations. Among the results, only marginal changes of the neutronics effects were observed. The T/H results, which included temperature and density of the fuel and the coolant, were not critical. Through this study, we are now able to determine in new ways, conventional derating values from a pressure tube.