Deformation analysis of fuel pins within the wire-wrap assembly of an LMFBR

Deformation of fuel pins within the wire-wrap fuel assembly of a fast breeder reactor is analyzed by two computational codes, the subchannel deformation analysis code SHADOW and the thermal-hydraulic analysis code DIANA. Coupling these codes makes it possible to analyze percisely the mechanical interactions between all fuel pins in an assembly, and the deviation of coolant temperature distribution in deformed flow channels from the nominal distribution.In this paper, particular attention is paid to the effect on fuel pin deformation of the following factors: dimensional changes in the fuel assembly components, displacement of wrapper tube walls and changes in the radial power gradients.     

Turbulent flow simulation in a wire-wrap rod bundle of an LMFBR

The pressure drop and heat transfer characteristics of wire-wrapped 19-pin rod bundles in a nuclear reactor subassembly of liquid metal cooled fast breeder reactor (LMFBR) have been investigated through three-dimensional turbulent flow simulations. The predicted results of eddy viscosity based turbulence models (k-?, k-ω) and the Reynolds stress model are compared with those of experimental correlations for friction factor and Nusselt number. The Re is varied between 50,000 and 150,000 and the ratio of helical pitch of wire wrap to the rod diameter is varied from 15 to 45. All the three turbulence models considered yield similar results. The friction factor increases with reduction in the wire-wrap pitch while the heat transfer coefficient remains almost unaltered. However, reduction in the wire-wrap pitch also enhances the transverse flow velocity in the cross-sectional plane as well as the local turbulence intensity, thereby improving the thermal mixing of coolant. Consequently, the presence of wire wrap reduces temperature variation within each section of the subassembly. The associated reduction in differential thermal expansion of rods is expected to improve the structural integrity of the fuel subassembly.     

Channel deformation analysis for fast reactor fuel assemblies undergoing swelling and thermal bowing

Differential thermal expansion and swelling of fuel pins, hexagonal flow ducts and fuel spin spacers, fuel pin bowing between the spacers due to subchannel temperature differentials, and fuel pin bundle bowing due to the cambered hexagonal wrapper tube were analyzed for sodium-cooled fast reactor fuel assemblies.     

Application of X-Ray Computer Tomography for Observing the Deflection and Displacement of Fuel Pins in an Assembly Irradiated in FBR

A nondestructive method making use of X-ray computer tomography (X-ray CT) has been applied to post irradiation examination of fast breeder reactor (FBR) fuel assemblies. In the study, an examination is made of the deflection and displacement of fuel pin in a fuel assembly irradiated to 74.2GWd/t peak burnup in the fast reactor “JOYO.”

In the examination, X-ray CT images of transverse cross sections of fuel pin were obtained at different heights of fuel pin along its axis. Analysis of the resulting images indicated that:

1. The hexagonal wrapper tube had its lateral wall faces slightly bulged outward;

2. The fuel pins loaded in the outermost array were markedly displaced in the direction of wrapper tube, particularly in portions of fuel pin intermediate between positions constrained by wrapping wire.

The latter behavior of fuel pins was substantiated by the contours of fuel pin along its axis, which were derived from cross section images obtained at different levels along axis.

Such fuel pin displacement is surmised to have been caused by thermal stressing of the affected fuel assembly cladding.     

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.     

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.     

CFD investigations of helical wire-wrap fuel pin bundle and its comparison with straight wire bundle

Computational Fluid Dynamics (CFD) investigations of a fast reactor fuel pin bundle wrapped with helical and straight spacer wires have been carried out and the advantages of using helical spacer wire have been assessed. The flow and temperature distributions in the fuel pin bundle are obtained by solving the statistically averaged 3-Dimensional conservation equations of mass, momentum and energy along with high Reynolds number k-ε turbulence model using a customized CFD code CFDEXPERT. It is seen that due to the helical wire-wrap spacer, the coolant sodium not only flows in axial direction in the fuel pin bundle but also in a transverse direction. This transverse flow enhances mixing of coolant among the sub channels and due to this, the friction factor and heat transfer coefficient of the coolant increase. Estimation of friction factor, Nusselt number, sodium temperature uniformity at the outlet of the bundle and clad hot spot factor which are measures of the extent of coolant mixing and non-homogeneity in heat transfer coefficient around fuel pin are paid critical attention. It is seen that the friction factor and Nusselt number are higher (by 25% and 15% respectively) for the helical wire wrap pin bundle compared to straight wire bundle. It is seen that for 217 fuel pin bundle the maximum clad temperature is 750 K for straight wire case and the same for helical wire is 720 K due to the presence of transverse flow. The maximum temperature occurs at the location of the gap between pin and wire. The ΔT between the bulk sodium in the central sub-channel and peripheral sub-channel is 30 K for straight wire and the same for helical wire is 18 K due to the presence of secondary transverse flow which makes the outlet temperature more uniform. The hotspot factor and the hot channel factors predicted by CFD simulation are 10% lower than that used in conventional safety analysis indicating the conservatism in the safety analysis.     

Analytical Method of Fuel Pin Deformations in LMFBR Assembly

A new analytical method is presented for analyzing, in three dimensions, the mechanical response of fuel pins with wire spacers, to their thermal and neutronic environment in a fuel assembly of a LMFBR. It analyzes the mechanical interactions between fuel pins in the assembly in each of three directions, which form an angle of π/3 radians with one another, based on the mathematical relationship between the displacements at the contact points and the associated contact forces with respect to all fuel pins forming a line in one of the three directions.

Based on this method, a new computational code, the Subchannel Deformation Analysis Code for Wire-Wrap Assemblies (SHADOW) has been developed, and is applied to a fuel assembly of a prototype fast breeder reactor in order to analyze the deformation of 169 fuel pins due to thermal bowing.

Conclusions drawn from the study confirm that the SHADOW code can be an effective tool for analyzing or evaluating thermal and structural designs of a LMFBR fuel assembly.     

A second order turbulence model based on a Reynolds stress approach for two-phase boiling flow. Part 1: Application to the ASU-annular channel case

High-thermal performance PWR (pressurized water reactor) spacer grids require both low pressure loss and high critical heat flux (CHF) properties. Numerical investigations on the effect of angles and position of mixing vanes and to understand in more details the main physical phenomena (wall boiling, entrainment of bubbles in the wakes, recondensation) are required.In the field of fuel assembly analysis or design by means of CFD codes, the overwhelming majority of the studies are carried out using two-equation eddy viscosity models (EVM), especially the standard K-? model, while the use of Reynolds Stress Transport Models (RSTM) remains exceptional.But extensive testing and application over the past three decades have revealed a number of shortcomings and deficiencies in eddy viscosity models. In fact, the K-? model is totally blind to rotation effects and the swirling flows can be regarded as a special case of fluid rotation. This aspect is crucial for the simulation of a hot channel in a fuel assembly. In fact, the mixing vanes of the spacer grids generate a swirl in the coolant water, to enhance the heat transfer from the rods to the coolant in the hot channels and to limit boiling.First, we started to evaluate computational fluid dynamics results against the AGATE-mixing experiment: single-phase liquid water tests, with Laser-Doppler liquid velocity measurements upstream and downstream of mixing blades. The comparison of computed and experimental azimuthal (circular component in a horizontal plane) liquid velocity downstream of a mixing vane for the AGATE-mixing test shows that the rotating flow is qualitatively well reproduced by CFD calculations but azimuthal liquid velocity is underestimated with the K-? model.Before comparing performance of EVM and RSTM models on fuel assembly geometry, we performed calculations with a simpler geometry, the ASU-annular channel case. A wall function model dedicated to boiling flows is also proposed.     

Benchmark Problem Suite for Reactor Physics Study of LWR Next Generation Fuels

This paper proposes a benchmark problem suite for studying the physics of next-generation fuels of light water reactors. The target discharge burnup of the next-generation fuel was set to 70GWd/t considering the increasing trend in discharge burnup of light water reactor fuels. The UO₂ and MOX fuels are included in the benchmark specifications. The benchmark problem consists of three different geometries: fuel pin cell, PWR fuel assembly and BWR fuel assembly. In the pin cell problem, detailed nuclear characteristics such as burnup dependence of nuclide-wise reactivity were included in the required calculation results to facilitate the study of reactor physics. In the assembly benchmark problems, important parameters for in-core fuel management such as local peaking factors and reactivity coefficients were included in the required results. The benchmark problems provide comprehensive test problems for next-generation light water reactor fuels with extended high burnup. Furthermore, since the pin cell, the PWR assembly and the BWR assembly problems are independent, analyses of the entire benchmark suite is not necessary: e.g., the set of pin cell and PWR fuel assembly problems will be suitable for those in charge of PWR in-core fuel management, and the set of pin cell and BWR fuel assembly problems for those in charge of BWR in-core fuel management.     

Numerical investigation of a heat transfer within the prismatic fuel assembly of a very high temperature reactor

The complex geometry of the hexagonal fuel blocks of the prismatic fuel assembly in a very high temperature reactor (VHTR) hinders accurate evaluations of the temperature profile within the fuel assembly without elaborate numerical calculations. Therefore, simplified models such as a unit cell model have been widely applied for the analyses and designs of prismatic VHTRs since they have been considered as effective approaches reducing the computational efforts. In a prismatic VHTR, however, the simplified models cannot consider a heat transfer within a fuel assembly as well as a coolant flow through a bypass gap between the fuel assemblies, which may significantly affect the maximum fuel temperature. In this paper, a three-dimensional computational fluid dynamics (CFD) analysis has been carried out on a typical fuel assembly of a prismatic VHTR. Thermal behaviours and heat transfer within the fuel assembly are intensively investigated using the CFD solutions. In addition, the accuracy of the unit cell approach is assessed against the CFD solutions. Two example situations are illustrated to demonstrate the deficiency of the unit cell model caused by neglecting the effects of the bypass gap flow and the radial power distribution within the fuel assembly.     

基于CFD方法的铅铋冷却燃料棒束的热工水力特性分析

铅铋冷却快堆作为第4代反应堆候选之一具有安全性高等特点,研究其在正常工况下的热工水力特性具有重要意义。本文基于商用计算流体力学（CFD）软件STAR-CCM+,使用流固耦合的方法对带有绕丝结构的19棒束铅铋组件进行数值分析,探究了质量流量、功率等边界条件对组件内部流动传热特性的影响。模拟计算结果表明：CFD方法在子通道中心温度和壁面温度预测上与实验结果取得了较好的一致。同时,绕丝结构的存在使得子通道之间存在周期性的横向交混,并使得棒束表面温度呈现震荡。随质量流量的增加,子通道间横向交混增大。功率变化对通道间的横向交混速度的影响较小,冷却剂温度的横向分布无明显差异。     

Thermal shock in a nuclear fuel element with cladding

The object of the present paper is to look into the evolution of thermoelastic state in a nuclear pin with cladding, when a step-variation occurs in coolant temperature.For both fuel and cladding the transient stress components are obtained: they must be added to the steady stress components in the case of step-decrease (ΔT < 0), and subtracted in the case of step-growth (ΔT > 0). Some numerical results are finally reported with reference to a fast breeder reactor cooled by sodium and to a pressurized water reactor.     

CFD analysis of heat transfer in subchannels of a Super Fast Reactor

This paper presents CFD analyses of heat transfer in subchannels of a Super Fast Reactor fuel assembly. Analyses are concentrated on the circumferential temperature distribution on the cladding outer surface because the Maximum Cladding Surface Temperature (MCST) has been a crucial design parameter to evaluate fuel cladding integrity of the Super Fast Reactor. Speziale non-linear high Re k-? model, which can reproduce the anisotropic turbulence flow in non-circular flow channels, with two-layer near-wall treatment is adopted. The results show that heat conduction in the cladding should be considered in the CFD analyses. Larger circumferential temperature gradient occurs on the cladding surface in the edge and corner subchannels than that in the ordinary subchannel because of their special geometries causing larger heterogeneity of mass flow rate distribution inside the subchannels. Improved subchannel configurations to reduce the circumferential temperature gradient are proposed. This study will be a good guideline to the future core design improvement.     

基于流固共轭传热的两相流动数值模拟及燃料子通道CHF预测研究

为对过冷沸腾两相流动进行准确模拟,并探索临界热流密度（CHF）预测方法,本文基于共轭传热和两相CFD分析的方法,通过流固界面耦合,建立流固共轭传热两相流动耦合求解的数值模型。首先通过典型燃料棒栅元过冷沸腾两相流动的模拟,验证数值模型的正确性。随后对燃料子通道内两相流动进行模拟,并在两相流动模拟的基础上,通过准瞬态的方法,建立与CHF试验过程非常近似的CHF预测方法,将加热壁面的温度飞升作为CHF判定的标准,实现对燃料组件子通道CHF的数值预测。研究表明,本文建立的数值模拟方法,可为燃料组件或其他换热系统的CHF预测奠定基础,为燃料组件的设计提供新的辅助手段。     

Study on Two-phase Flow in Fluid Channel of Lead-based Fast Reactor Based on OpenFOAM

In order to study the thermal-hydraulic characteristics of two-phase flow caused by the special thermal properties of lead/lead-bismuth in lead-based fast reactors, the influence of bubble in fluid channel on the heat transfer capacity and safety of the core was simulated. In this paper the open source CFD calculation software OpenFOAM was adopted, and the numerical simulation was applied based on VOF method to construct a common triangular channel model in lead-based fast reactor. By simulating the two-phase flow of the coolant channel, it is found that as the flow rate increases, the outlet temperature of the coolant decreases. In the flow process of the gas-liquid two-phase flow in the channel, it can be found that the gas phase basically flows inside the channel. In the simulation of the fuel assembly, the corner channel is an area with a large amount of bubbles, which will cause the local heat transfer to deteriorate and cause the fuel assembly to burn out.     

基于CFD程序的物理热工耦合计算不确定性分析

基于计算流体力学（CFD）程序FLUENT的用户自定义函数（UDF）,耦合中子动力学计算模型、燃料棒热传导计算模型、不确定性分析程序SIMLAB,开发了物理热工耦合计算不确定性分析平台CFD/PFS,并开展了小型自然循环铅基快堆SNCLFR-10的无保护超功率（UTOP）事故的不确定性量化,最后对计算结果进行不确定性分析和敏感性分析。研究表明,CFD/PFS平台的物理热工耦合计算具有良好的可靠性、精确性;总反应性峰值、功率峰值等瞬态安全参数的名义值均处于95/95双侧容忍限值内,且名义值与限值相对偏差小于3.95%;燃料多普勒系数是主要不确定性来源,对反应堆安全影响最大。     

基于OpenFOAM的铅基快堆流体通道内两相流研究

为研究铅基快堆中铅/铅铋的特殊热物性导致的在两相流情况下的热工水力特性,模拟流体通道中空泡存在对堆芯的输热能力以及安全性的影响,本文采用开源的CFD计算软件OpenFOAM,应用基于VOF方法的数值模拟,构建了铅基快堆中常见的三角形通道模型,通过与子通道程序的验证和单相条件下实验的校核,检验了所用代码的准确性,并对堆内冷却剂通道的两相流进行了模拟。模拟结果表明：随着两相流流速的增大,冷却剂出口温度降低。气液两相流在内通道流动过程中,气相基本在通道内部流动。随着轴向高度的升高,气泡会在内通道的中心区域聚合;燃料组件的角通道是气泡含量多的区域,会造成局部传热恶化,导致组件烧毁。     

Development of supercritical water heat-transfer correlation for vertical bare tubes

This paper presents an analysis of heat-transfer to supercritical water in bare vertical tubes. A large set of experimental data, obtained in Russia, was analyzed and a new heat-transfer correlation for supercritical water was developed. This experimental dataset was obtained within conditions similar to those in supercritical water-cooled nuclear reactor (SCWR) concepts.The experimental dataset was obtained in supercritical water flowing upward in a 4-m long vertical bare tube with 10-mm ID. The data were collected at pressures of about 24 MPa, inlet temperatures from 320 to 350 °C, values of mass flux ranged from 200 to 1500 kg/m² s and heat fluxes up to 1250 kW/m² for several combinations of wall and bulk-fluid temperatures that were below, at, or above the pseudocritical temperature.A dimensional analysis was conducted using the Buckingham Π-theorem to derive the general form of an empirical supercritical water heat-transfer correlation for the Nusselt number, which was finalized based on the experimental data obtained at the normal and improved heat-transfer regimes. Also, experimental heat transfer coefficient (HTC) values at the normal and improved heat-transfer regimes were compared with those calculated according to several correlations from the open literature, with CFD code and with those of the proposed correlation.The comparison showed that the Dittus-Boelter correlation significantly overestimates experimental HTC values within the pseudocritical range. The Bishop et al. and Jackson correlations tended also to deviate substantially from the experimental data within the pseudocritical range. The Swenson et al. correlation provided a better fit for the experimental data than the previous three correlations at low mass flux (∼500 kg/m² s), but tends to overpredict the experimental data within the entrance region and does not follow up closely the experimental data at higher mass fluxes. Also, HTC and wall temperature values calculated with the FLUENT CFD code might deviate significantly from the experimental data, for example, the k-? model (wall function). However, the k-? model (low Reynolds numbers) shows better fit within some flow conditions.Nevertheless, the proposed correlation showed the best fit for the experimental data within a wide range of flow conditions. This correlation has an uncertainty of about ±25% for calculated HTC values and about ±15% for calculated wall temperature. A final verification of the proposed correlation was conducted through a comparison with other datasets. It was determined that the proposed correlation closely represents the experimental data and follows trends closely, even within the pseudocritical range. Finally, a recent study determined that in the supercritical region, the proposed correlation showed the best prediction of the data for all three sub-regions investigated.Therefore, the proposed correlation can be used for HTC calculations in SCW heat exchangers, for preliminary HTC calculations in SCWR fuel bundles as a conservative approach, for future comparison with other datasets and for the verification of computer codes and scaling parameters between water and modelling fluids.