Multi-dimensional analysis of two-phase flow inside subchannels of nuclear fuel assembly

In subchannel analysis, the conservation equations are solved for each channel in a complex fuel bundle, where the effects of fluid exchange between each subchannel are considered. The fluid exchange is commonly referred to as that caused by cross flow. Void drift is considered to be phenomenon resulting from attaining a hydrodynamic equilibrium state. Its mechanism has not been clarified, and the transport due to void drift is therefore estimated through empirical models in conventional subchannel analyses. Therefore, mechanistic model for the void drift phenomenon is required to apply the subchannel analysis to a variety of fuel bundle geometry. In this study, multi-dimensional analysis using two-fluid model was applied to two-phase flow inside a geometry simulating fuel bundle subchannels, for the purpose of clarifying the void drift mechanism. The comparison between the results of the numerical analysis and the experiment confirmed that the reliability of the numerical method used in this study. In this paper, a mechanistic model based on the Stanton number, which expresses the void diffusion coefficient based on the Lahey's proposal, was proposed.     

Analysis of low Reynolds number turbulent flow phenomena in nuclear fuel pin subassemblies of tight lattice configuration

This paper focuses on the numerical simulation of low Reynolds (Re) number turbulence flow phenomena in tightly packed fuel pin subassemblies and in channels of irregular shape such as eccentric annuli. Highlighted phenomena include (i) turbulence-driven secondary flows inside a subchannel, (ii) local turbulent-laminar transition in the narrow gap region, and (iii) global flow pulsation across the gap along the channel length. These phenomena are simulated by Computational Fluid Dynamics (CFD). The CFD methods employed here are those of Direct Numerical Simulation (DNS) of turbulence, Large Eddy Simulation (LES) and Reynolds-Averaged Navier-Stokes (RANS) equations approach. Complicated turbulent flow structure is due to strong anisotropy in the non-uniform channel geometry that is characterized by wide open channels connected by a narrow gap. The secondary flows in subchannels play an important role in transporting small eddies generated in the wider region toward the narrow gap. Periodic cross-flow oscillations are calculated to appear in the vicinity of the gap region, and the coherent structure is transported in the main flow direction. This macroscopic flow process prevails in the low Re turbulent flow regime and is called as global flow pulsation. Finally a brief discussion is made on their influences onto the mixing between two subchannels that must be taken into account during natural circulation decay heat removals.     

Single-phase subchannel mixing in a simulated nuclear fuel assembly

Single-phase subchannel mixing data were obtained from a 25 rod square array by measuring precisely subchannel exit temperatures over a range of test conditions. A least-squares type statistic operating on exit enthalpy differences was developed to ascertain, in conjunction with the COBRA-II subchannel computer analysis, an optimum value for the coefficient of Rowe and Angles's single-phase mixing correlation β. When the same analytical procedures were applied to data taken under conditions of subcooled nucleate boiling, an approximately linear correlation of β with average exit quality was found. The values obtained were for conditions of natural, or unaided mixing, and for design purposes should be considered as a dependable lower limit. Commercial rod-spacer designs intended to increase mixing will, of course, show substantially higher values of β.     

板型燃料组件额定流速流致振动实验研究

针对贫铀叠层板型燃料组件,采用了一种新的软测量方法,进行了其在经受冷却剂额定流速冲刷时的振动实验。获得了该组件在额定流速6 m/s下的流致振动动态响应时间域和频率域结果,并对实验结果进行了分析。     

Optimization of mine layout for nuclear fuel assembly storage

The development of a mine layout for a nuclear waste repository in bedded salt is discussed. Optimizations of storage room arrangements and mine arrangements are addressed separately, and a conceptual repository design based on these optimizations is described.     

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.     

Experimental investigation of turbulent flow through spacer grids in fuel rod bundles

This paper contains experimental data of pressure, velocity and turbulence intensity in a 24-rod fuel bundle with spacer grids. Detailed pressure measurements inside the spacer grid have been obtained by use of a sliding pressure-sensing rod. Laser Doppler Velocimetry technique was used to measure the local axial velocity and its fluctuating component upstream and downstream of the spacer grid in sub-channels with different blockage ratios. The measurements show a changing pattern in function of radial position in the cross-section of the fuel bundle. For sub-channels close to the box wall, the turbulence intensity suddenly increases just downstream of the spacer and then gradually decays. In inner sub-channels, however, the turbulence intensity downstream of the spacer decreases below its upstream value and then gradually increases until it reaches the maximum value at approximately two spacer heights. The present study reveals that spacer effects, such as local pressure distribution and turbulence intensity enhancement, not only depend exclusively on the local geometry details, but also on the location in the cross-section of the rod bundle.     

Sintering of oxide nuclear fuel: Plastic flow mechanism

A theoretical approach to solving the sintering problem of nuclear oxide fuels is described. We are proposing a new physical sintering model: the main mechanism for oxide nuclear shrinkage is the plastic flow of the matter under internal stress and capillary forces.     

Turbulent flow in a model nuclear fuel rod bundle containing partial flow blockages

Local velocity and turbulence intensity measurements were obtained with a laser Doppler anemometer near flow blockages in an unheated 7 × 7 rod bundle. Sleeve blockages were positioned on the center nine rods to create area reductions of 70 and 90% in the center four subchannels of the bundle. Experimental results indicated that extensive flow disturbances existed downstream from the blockage clusters and showed that only minor disturbances can be expected upstream from the blockages. Recirculation zones for both 70 and 90% blockages were detected downstream from the blockage clusters and persisted for approximately three to five subchannel hydraulic diameters, depending on the degree of the blockage. The experimental velocity results obtained with blockage clusters located midway between grid spacers were successfully predicted using the COBRA subchannel computer program.     

Large eddy simulation of turbulent flow surrounding two simulated CANDU fuel bundles

Three-dimensional turbulent flow in one and a half simulated 43-element CANDU^®1 fuel bundle at the inlet of a fuel channel is solved using large eddy simulation. Wake generated after the endplates and the flow development in the inlet bundle are investigated based on the simulation results. Spatial distribution and frequency spectra of fluid force components are also examined. The simulation results are compared to experimental data available in the literature as well as the measurements conducted by the authors. The current investigation provides a basic understanding of the fluid excitation in a simulated 43-element fuel bundle. The results may be used in a flow-induced vibration analysis for fuel bundles.     

CFD study on coolant mixing in VVER-440 fuel rod bundles and fuel assembly heads

A CFD model of VVER-440 fuel assembly heads was developed based on the technical documentation of a full-scale test facility built in the Kurchatov Institute, Russia. Steady-state and transient calculations were performed to validate the model with a measurement set. Effects of the spatial resolution, turbulence models, difference schemes and different inlet boundary conditions were investigated. Inlet boundary conditions were determined with both the COBRA subchannel code and a fuel rod bundle CFD model that was built for this special purpose. The results were compared against experimental data. The sensitivity studies showed that a grid of about 8 million cells, high resolution scheme and BSL Reynolds stress model are suitable sets to provide accurate prediction for the signal of the in-core thermocouple. The best prediction was achieved with transient calculation using inlet boundary conditions generated with the CFD fuel rod bundle model. The results indicated that the coolant mixing is intensive but not perfect in the assembly head. Besides, the significant role of the outflow from the central tube was also proven. The transient runs revealed relatively large temperature fluctuations near the in-core thermocouple housing.     

Partial flow blockage analysis of the hottest fuel assembly in SNCLFR-100 reactor core

In this paper, we perform an unprotected partial flow blockage analysis of the hottest fuel assembly in the core of the SNCLFR-100 reactor, a 100 MW_(th) modular natural circulation lead-cooled fast reactor, developed by University of Science and Technology of China. The flow blockage shall cause a degradation of the heat transfer between the fuel assembly and the coolant potentially,which can eventually result in the clad fusion. An analysis of core blockage accidents in a single assembly is of great significance for LFR. Such scenarios are investigated by using the best estimation code RELAP5. Reactivity feedback and axial power profile are considered. The crosssectional fraction of blockage, axial position of blockage,and blockage-developing time are discussed. The cladding material failure shall be the biggest challenge and shall be a considerable threat for integrity of the fuel assembly if the cross-sectional fraction of blockage is over 94%. The blockage-developing time only affects the accident progress. The consequence will be more serious if the axial position of a sudden blockage is closer to the core outlet.The method of analysis procedure can also be applied to analyze similar transient behaviors of other fuel-type reactors.     

Flooding in a thin rectangular slit geometry representative of ATR fuel assembly side-plate flow channels

Air-water counter-current flow limit experiments were conducted in thin rectangular channels at atmospheric pressure. The parameters were: narrow channel width, either 1.1 mm or 2.2 mm; inlet water temperature, ranging from 294 K to 330 K; channel surface condition, either clean aluminum, aluminum oxide, or acrylic; location and geometry of the air inlet; method of forcing air through the channel; and liquid head above the channel. Experimental results for each set of parameters can be linearly correlated using the square root of the non-dimensional superficial velocities. Channel surface wetting and location and geometry of the air injection had the greatest effect. Narrow channel width, water temperature, method used to force air through the channel, and liquid head above the channel had little effect on the flooding characteristics.     

A study of acoustic wave damping in water-filled pipes with zero flow and turbulent flow

A simple, fundamental experimental study was conducted to further the understanding of acoustic wave propagation in fluid-filled pipes. Three experiments were conducted: the first with zero flow and a closed outlet end; the second with turbulent flow and an open outlet end; and the third with zero flow and an open outlet end. The intent was to obtain data at higher frequencies than those previously reported and which can be used to validate and verify numerical models. A further objective was to determine the effect of turbulent flow on the acoustic response of the system. Some new insights are obtained and presented.     

A three-dimensional study of nuclear fuel rod behavior during startup

A three-dimensional finite element study is made of the behavior of cylindrical uranium dioxide fuel pellets during startup. The finite element code uses an eight-noded box element of arbitrary shape to build up the stiffness and stress characteristics by Gaussian integration. Each box has 33 degrees of freedom: 24 corresponding to the three motions at each of the eight nodes; and nine internally eliminated to minimize strain energy. The nine internal degrees of freedom are highly effective in eliminating shear error, and thus permitting far fewer elements than are required when tetrahedron elements are used. The element uses an isoparametric approach, so that the box can have eight arbitrarily positioned nodes. So long as the thermal expansions of the fuel rod can be described by a linear variation in the element, the code takes highly accurate account of it. Plasticity is accounted for by the secant modulus approach. Friction between the pellet and the cladding can be introduced by springs between the relevant finite elements in each area. A feature of the analysis allows cracks to appear in the uranium dioxide fuel as it is heated and the growth of the cracks can be traced as a function of linear power generated in the rod. The code can predict such things as pellet deformations and the stress and strain distributions within the pellets and the cladding. The three-dimensionality of the analysis allows a detailed view of these stresses and strains, and the interaction between the axial and plane stress distributions.     

Feasibility study of the application of exotic pin for tight-lattice fuel assembly

A tight-lattice fuel assembly having less space for the coolant is more feasibly applied in Liquid Metal Fast Breeder Reactor (LMFBR). The thermal hydraulic constraint due to smaller coolant space can be compensated by the high heat capacity of the liquid metal coolant. A tight pin configuration provides high fuel volume fraction which eventually gives better neutronic performance for longer core lifetime. A cylindrical pin array provides less flexible arrangement for tight-lattice assembly, which results in very narrow coolant gaps connecting its neighboring subchannels. Therefore, the so-called exotic pin shape is introduced, which enable to distribute the coolant flow more uniformly, to be applied in tight-lattice bundles with sodium coolant. As Nusselt number and wall friction correlation are absent for this type of geometry, CFD calculations are performed by employing k-ε turbulent model.     

秦山核电厂反应堆燃料组件的流量平衡试验研究

本文描述了对我国第一台自己设计建造的三十万千瓦压水堆核电站采用的燃料组件之间的流量平衡进行了一系列的水力特性试验研究,合理地解决了燃料组件上方四种不同结构部件间的阻力匹配,把燃料组件之间的冷却剂流量偏差调整在1％以内。同时,通过实验改进了阻力塞部件的结构设计,确定了反应堆堆芯上栅格板的开孔尺寸,测定了各种不同形式燃料组件的出口阻力系数,为秦山核电厂反应堆的热工设计和结构设计提供了可靠的实验依据。