CFD prediction of flow and phase distribution in fuel assemblies with spacers

This paper is concerned with the modeling and computation of multi-dimensional two-phase flows in BWR fuel assemblies. The modeling principles are presented based on using a two-fluid model in which lateral interfacial effects are accounted for. This model has been used to evaluate the velocity fields of both vapor and liquid phases, as well as phase distribution, between fuel elements in geometries similar to BWR fuel bundles. Furthermore, this model has been used to predict, in a detailed mechanistic manner, the effects of spacers on flow and phase distribution between, and pressure drop along, fuel elements. The related numerical simulations have been performed using a CFD computer code, CFDS-FLOW3D.     

A CFD simulation process for fast reactor fuel assemblies

A CFD modeling and simulation process for large-scale problems using an arbitrary fast reactor fuel assembly design was evaluated. Three-dimensional flow distributions of sodium for several fast reactor fuel assembly pin spacing configurations were simulated on high performance computers using commercial CFD software. This research focused on 19-pin fuel assembly “benchmark” geometry, similar in design to the Advanced Burner Test Reactor, where each pin is separated by helical wire-wrap spacers. Several two-equation turbulence models including the k-? and SST (Menter) k-ω were evaluated. Considerable effort was taken to resolve the momentum boundary layer, so as to eliminate the need for wall functions and reduce computational uncertainty. High performance computers were required to generate the hybrid meshes needed to predict secondary flows created by the wire-wrap spacers; computational meshes ranging from 65 to 85 million elements were common. A general validation methodology was followed, including mesh refinement and comparison of numerical results with empirical correlations. Predictions for velocity, temperature, and pressure distribution are shown. The uncertainty of numerical models, importance of high fidelity experimental data, and the challenges associated with simulating and validating large production-type problems are presented.     

Thermal analysis of a storage cask for 24 spent PWR fuel assemblies

Abstract

The purpose of this paper is to perform a thermal analysis of a spent fuel storage cask in order to predict the maximum concrete and fuel cladding temperatures. Thermal analyses have been carried out for a storage cask under normal, off-normal and accident conditions. The environmental temperature is assumed to be 27°C under the normal condition. The off-normal condition has an environmental temperature of 40°C. An additional off-normal condition is considered as a partial blockage of the air inlet ducts. Four of the eight inlet ducts are assumed to be completely blocked. The accident condition is defined as a 100% blockage of air inlet ducts. The storage cask is designed to store 24 PWR spent fuel assemblies with a burn-up of 55,000 MWD/MTU and a cooling time of 7 years. The decay heat load from the 24 PWR assemblies is 25.2 kW. Thermal analyses of the ventilation system have been carried out for the determination of the optimum duct size and shape. The finite-volume computational fluid dynamics code FLUENT was used for the thermal analysis. From the results of the analysis, the maximum temperatures of the fuel rod and concrete overpack were lower than the allowable values under the normal, off-normal and accident conditions.     

On the factors affecting the fretting-wear risk of PWR fuel assemblies

Key factors affecting the rod-to-grid fretting-wear risk of fuel assemblies operated in pressurized water reactors (PWR) are evaluated. The analysis is part of a comprehensive approach to predict fretting-wear risk based on the fuel assembly operating conditions. The assembly wear damage is determined by a non-linear vibration model of the nuclear fuel rod exposed to a turbulent flow. The study evaluates the sensitivity of the wear damage to the grid support forces, fuel rod-to-grid gap size, assembly grids misalignment, rod structural damping and stiffness, assembly bow shape, friction coefficients and turbulence force spectrum. The results of the numerical simulations show that the grid cell clearance and the turbulence forces are key factors in the wear process. Since a good correlation exists between these two parameters and the assembly location in the core, it is recommended to include consideration of the wear risk minimization as an additional criterion for the design of the core loading pattern.     

Enthalpy transfer between PWR fuel assemblies in analysis by the lumped subchannel model

Assessment is made of the error in enthalpy predictions in PWR cores resulting from various assumptions in the size of homogenized calculation regions. A coefficient is developed from the differential conservation equations and numerically determined to minimize these errors. It was found that these errors are only of significance for regions comprising three to seven subchannels, region sizes which may be employed in coupled neutronic-thermal hydraulic calculations. A correlation for coefficient values for a range of PWR conditions is presented.     

The study on a statistical methodology for PWR fuel rod internal pressure evaluation

The most limiting design criteria for high Burnup PWR fuel are known to be rod internal pressure and cladding oxidation. Some fuel vendors have been increasing the design margin of rod internal pressure by increasing fuel rod plenum volume or optimizing fuel pellet grain size. In this study, a sophisticated statistical methodology that employs the response surface method and Monte Carlo simulation has been proposed to increase the design margin of rod internal pressure and subsequently a simplified statistical methodology has been developed to simplify the sophisticated statistical methodology. The simplified statistical methodology utilizes the system moment method combined with a deterministic approach for calculating a maximum variance of rod internal pressure. This simplified statistical methodology may be more efficient in the reload core fuel rod performance analyses than the sophisticated statistical methodology since the former eliminates numerous calculations needed for the evaluation of power history-dependent variances. It is found that this simplified methodology also generates more conservative rod internal pressure than the typical statistical methodology.     

Super-cell model for predicting local power distribution in BWR type fuel assemblies and its experimental validation

A super-cell model is described for the prediction of the local power distribution in BWR type fuel assemblies. The model is validated against the measured power distribution in nine fuel-loading patterns reported in the literature. Overall agreement is observed to be quite satisfactory. Local Peaking Factor (LPF) is predicted within a maximum uncertainity of 3.7%. The maximum root-mean-square error among the nine fuel loading patterns is observed to be 2.3%.     

RANS modeling for flow in nuclear fuel bundle in pressurized water reactors (PWR)

This paper presents use of Reynolds-averaged Navier-Stokes (RANS) based turbulence model for single-phase CFD analysis of flow in pressurized water reactor (PWR) assemblies. An open source code called OpenFoam was used for computational fluid dynamics (CFD) study using computational meshes generated using Shari Harpoon. The PWR assembly design used in this analysis represents a 5 × 5 pin design including structural grid equipped with mixing vanes. The design specifications used in this study were obtained from the experimental setup at Texas A&M University and the results obtained are used to validate the CFD software, algorithm, and the turbulence model used in this analysis.     

Row of fuel assemblies analysis under seismic loading: Modelling and experimental validation

The aim of this study was to develop a numerical model for predicting the impact behaviour at fuel assembly level of a whole reactor core under seismic loading conditions. This model was based on a porous medium approach accounting for the dynamics of both the fluid and structure, which interact. The fluid is studied in the whole reactor core domain and each fuel assembly is modelled in the form of a deformable porous medium with a nonlinear constitutive law. The contact between fuel assemblies is modelled in the form of elastic stops, so that the impact forces can be assessed. Simulations were performed to predict the dynamics of a six fuel assemblies row immersed in stagnant water and the whole apparatus was placed on a shaking table mimicking seismic loading conditions. The maximum values of the impact forces predicted by the model were in good agreement with the experimental data. A Proper Orthogonal Decomposition analysis was performed on the numerical data to analyse the mechanical behaviour of the fluid and structure more closely.     

Heat removal from VVÉR fuel assemblies for low mass flow rates

Translated from Atomnaya Énergiya, Vol. 68, No. 4, pp. 288–291, April, 1990.     

Heat transfer in turbulent longitudinal flow through unbaffled assemblies of fuel rods

A numerical analysis of heat transfer in turbulent longitudinal flow through assemblies of unbaffled fuel rods is presented. The solution applies to triangular or rectangular arrays of fuel rods with fully developed velocity and temperature profiles, for fluids with Prandtl number 1 and « 1. In the case of liquid metals, the thermal resistance of the cladding and bond are considered, but the turbulent heat transport component is neglected. For common liquids the circumferential turbulent heat transfer is considered. Results are compared in the range of dimensionless rod spacing of 1.0–1.6. Theoretical predictions and experimental results of other authors dealing with the problem show relatively good agreement.     

Numerically simulating the thermal-hydraulic characteristics within the fuel rod bundle using CFD methodology

With the dramatic progress in the computer processing power, computational fluid dynamics (CFD) methodology can be applied in investigating the detailed knowledge of thermal-hydraulic characteristics in the rod bundle, especially with the spacer grid. These localized information, including flow, turbulence, and heat transfer characteristics, etc., can assist in the design and the improvement of rod bundles for nuclear power plants. In this paper, a three-dimensional (3D) CFD model with the Reynolds stresses turbulence model is proposed to simulate these characteristics within the rod bundle and subsequently to investigate the effects of different types of grid on the turbulent mixing and heat transfer enhancement. Two types of grid designs are used herein, including the standard grid and split-vane pair one, respectively. Based on the CFD simulations, the secondary flow can be reasonably captured in the rod bundle with the grid. The split-vane pair grid would enhance both the flow mixing and the heat transfer capability more than the standard grid does, as clearly shown in the simulation results. In addition, compared with the results of experiment and correlation, the present predicted result for the Nusselt (Nu) number distribution downstream the grid shows reasonable agreement for the standard grid design. However, there is discrepancy in the decay trend of Nu number between the prediction and measurement for the split-vane pair gird. This would be improved by adopting the finer mesh (y⁺ < 1) simulation and Low-Reynolds form turbulence model, which is our future research work.     

A stepwise development and validation of a RANS based CFD modelling approach for the hydraulic and thermal-hydraulic analyses of liquid metal flow in a fuel assembly

The demand of reliable and clean energy at affordable prices poses a formidable challenge to the world. The initiatives from various international organizations reserve an important role for liquid metal cooled reactor systems. Assessment of such reactors usually involves unconventional thermal-hydraulics. Consequently, Reynolds Averaged Navier Stokes (RANS) based Computational Fluid Dynamics (CFD) approaches are expected to play a vital role along with various ongoing experiments for the design and the safe operation of these nuclear reactors. One of the major issues is the heat transport in the fuel assembly by the liquid metal. The known difficulties in heat transfer experiments, especially with liquid metals, necessitate the application of RANS in computing details of flow and temperature distribution.Considering these aspects, a four step approach is described in the current paper. As a first step, the heat transfer in a liquid metal flow inside a heated tube is analyzed using a RANS approach and then compared with some of the empirical correlations. The computed Nusselt number reveals the required development length of the thermal boundary layer in liquid metal. Furthermore, these simulations reveal the need of further assessment of this approach and all the existing correlations, and the care that should be taken while applying one of these correlations. In the second step, numerical simulations of the flow of heavy liquid metal around a heated rod in an annular cavity confirm that a RANS strategy can be employed in liquid metal flows. Furthermore, a comparison between computed and experimental non-dimensional axial temperature at the heated rod surface shows that among the considered turbulence models the use of a Baseline-Reynolds Stress Model (BSL-RSM) with automatic wall treatment (AWT) can be preferred for complex geometries. This is also demonstrated in the third step by computing the flow distribution in a triangular arrangement of a fuel assembly and by comparing with an existing hydraulics experiment in rod-bundle. These analyses reveal that the use of the BSL-RSM turbulence model with AWT allows prediction of the cross-flow in this rod-bundle. As the forth and last step, the integral TEGENA (TEmperatur- und GEschwindigkeitsverteilungen in Stabbündel mit turbulenter NAtriumströmung) experiment has been selected for further assessment of RANS based CFD approach in computing both the flow and temperature field. A comparison with literature demonstrates that the use of symmetric boundary condition in such a tightly packed parallel rod-bundle leads to a distorted flow field. The experimentally and the computationally obtained temperature field at a plane in the outlet reveal its acceptable predictive capability. Furthermore, application of two different Reynolds Stress Models yields almost the same temperature distribution as a result of the use of simple first-order gradient model for the turbulent heat fluxes. Consequently, these four steps support the use of this modelling approach for investigating the heat transport in (heavy) liquid metals. Finally, the preferred RANS approach has been applied for thermal-hydraulic evaluation in the square arrangement of a bare rod-bundle as is to be employed in the European Lead-cooled reactor System (ELSY). Also, the influence of rod pitch-to-diameter ratio has been analyzed by numerically re-arranging these rods in the different square lattice. Moreover, arranging the bare rods in triangular lattice at the different rod pitch-to-diameter ratios shows the effect of the square and triangular lattice in thermal-hydraulics. Lastly, comparisons with some of the existing heat transfer correlations for the triangular and the rectangular lattice allows us to identify preferred correlations for these lattice arrangements of bare and liquid metal cooled rod-bundles.     

CSNI validation matrix for PWR and BWR thermal-hydraulic system codes

An internationally agreed validation matrix for PWR and BWR thermal-hydraulic system codes has been established by the CSNI-PWG-2 Task Group on Status and Assessment of Codes for Transients and ECCS. The matrix will be a guide for independent code assessment, will be a basis for the comparisons of code predictions performed with different system codes, and may contribute to the quantification of the uncertainty range of code predictions.     

Experimental validation of radial reconstructed pin-power distributions in full-scale BWR fuel assemblies with and without control blade

Total fission rate measurements have been performed on full-size BWR fuel assemblies of type SVEA-96+ in the zero power reactor PROTEUS at the Paul Scherrer Institute. This paper presents comparisons of reconstructed 2D pin fission rates from nodal diffusion calculations to the experimental results in two configurations: one “regular” (I-1A) and the other “controlled” (I-2A). Both configurations consist of an array of 3 × 3 SVEA-96+ fuel assemblies moderated with light water at 20 °C. In configuration I-2A, an L-shaped hafnium control blade (half of a real cruciform blade) is inserted adjacent to the north-west corner of the central fuel assembly. To minimise the impact of the surroundings, all measurements were done in fuel pins belonging to the central assembly. The 3 × 3 experimental configuration (test zone) was modelled using the core monitoring and design tools that are applied at the Leibstadt Nuclear Power Plant (KKL). These are the 2D transport code HELIOS, used for the cross-section generation, and the 3D, 2-group nodal diffusion code PRESTO-2. The exterior is represented, in the axial and radial directions, by 2-group partial current ratios (PCRs) calculated at the test zone boundary using a 3D Monte Carlo (MCNPX) model of the whole PROTEUS reactor. Sensitivity cases are analysed to show the impact of changes in the 2D lattice modelling on the calculated fission rate distribution and reactivity. Further, the effects of variations in the test zone boundary PCRs and their behaviour in energy are investigated.     

Modeling and validation of a mechanistic tool (MEFISTO) for the prediction of critical power in BWR fuel assemblies

Westinghouse is currently developing the MEFISTO code with the main goal to achieve fast, robust, practical and reliable prediction of steady-state dryout Critical Power in Boiling Water Reactor (BWR) fuel bundle based on a mechanistic approach. A computationally efficient simulation scheme was used to achieve this goal, where the code resolves all relevant field (drop, steam and multi-film) mass balance equations, within the annular flow region, at the sub-channel level while relying on a fast and robust two-phase (liquid/steam) sub-channel solution to provide the cross-flow information. The MEFISTO code can hence provide highly detailed solution of the multi-film flow in BWR fuel bundle while enhancing flexibility and reducing the computer time by an order of magnitude as compared to a standard three-field sub-channel analysis approach.Models for the numerical computation of the one-dimensional field flowrate distributions in an open channel (e.g. a sub-channel), including the numerical treatment of field cross-flows, part-length rods, spacers grids and post-dryout conditions are presented in this paper. The MEFISTO code is then applied to dryout prediction in BWR fuel bundle using VIPRE-W as a fast and robust two-phase sub-channel driver code. The dryout power is numerically predicted by iterating on the bundle power so that the minimum film flowrate in the bundle reaches the dryout criteria. Predicted dryout powers (including trends with flow, pressure, inlet subcooling and power distribution) and predicted dryout locations (both axial and radial) are compared to experimental results, using the entire Westinghouse SVEA-96 Optima3 dryout database, and are shown to yield excellent results.