Study of Mixing Characteristics Over a Spacer Grid With Sloping Channels Based on Numerical Simulations in a PWR 5×5 Rod Bundle Using CFD Codes

2013 ◽  
Author(s):  
Xi Chen ◽  
Hong Zhang
Keyword(s):  
Numerical Simulations ◽  
Safety Margin ◽  
Local Heat Transfer ◽  
Local Heat ◽  
Spacer Grid ◽  
Pressurized Water ◽  
Spacer Grids ◽  
Rod Bundle ◽  
Commercial Code ◽  
Water Reactors

Spacer grids are important components of fuel assemblies for Pressurized Water Reactors (PWR). The presence of spacer grid promotes local heat transfer adjacent to the rod wall downstream by inducing swirl and cross flows within and between sub-channels to increase thermal hydraulic safety margin. Recent years, Computational Fluid Dynamics (CFD) methodologies are widely adopted to designs of spacer grids. This paper presents results of numerical simulations with commercial code CFX 12.0 in a PWR 5 × 5 rod bundle including a spacer grid with sloping channels. Based on a combined mesh generation approach of structured and unstructured mesh, distributions of velocity fields, temperature and pressure fields downstream the spacer grid were analyzed. The results indicate that cross flows caused by the spacer grid are uniform in circumference inducing no thermal hydraulic deterioration, but mass exchange between central hot fluid and external cold fluid appears insufficient for the new style grid.

Numerical Flow Simulation of Spacer Grids With Mixing Vanes in a 5 × 5 PWR Rod Bundle

2009 ◽  
Author(s):  
Moyse´s Alberto Navarro ◽  
Andre´ Augusto Campagnole dos Santos
Keyword(s):  
Heat Transfer ◽  
Pressure Drop ◽  
Pressure Loss ◽  
Considerable Increase ◽  
Local Heat Transfer ◽  
Local Heat ◽  
Spacer Grid ◽  
Spacer Grids ◽  
Rod Bundle ◽  
Vane Angle

The spacer grids exert great influence on the thermal hydraulic performance of the PWR fuel assembly. The presence of the spacers has two antagonistic effects on the core: an increase of pressure drop due to constriction on the coolant flow area and increase of the local heat transfer downstream the grids caused by enhanced coolant mixing. The mixing vanes, present in most of the spacer grid designs, cause a cross and swirl flow between and in the subchannels, enhancing even more the local heat transfer at the cost of more pressure loss. Due to this important hydrodynamic feature the spacer grids are often improved aiming to obtain an optimal commitment between pressure drop and enhanced heat transfer. In the present work, the fluid dynamic performance downstream a 5 × 5 rod bundle with spacer grids is analyzed with a commercial CFD code (CFX 11.0). Eleven different split vane spacer grids with angles from 16° to 36° and a spacer without vanes were evaluated. The computational domain extends from ∼10 Dh upstream to ∼50 Dh downstream the spacer grids. The standard k-ε turbulence model with scalable wall functions and the total energy model were used in the simulations. The results show a considerable increase of the average Nusselt number and secondary mixing with the angle of the vane up to ∼20 Dh downstream the spacer, reducing greatly the influence of the vane angle beyond this region. As expected, the pressure loss through the spacer grid also showed considerable increase with the vane angle.

Effect of Spacer Grids on the Local Heat Transfer

2010 ◽  
Author(s):  
Benediktas Cesna ◽  
Eugenijus Uspuras
Keyword(s):  
Heat Transfer ◽  
Experimental Data ◽  
Boundary Layer ◽  
Local Heat Transfer ◽  
Local Heat ◽  
Elementary Cell ◽  
Spacer Grid ◽  
Transverse Mixing ◽  
Spacer Grids ◽  
Fuel Rod

The paper presents the experimental data of the local heat transfer in the fuel rod assembly with honeycomb type spacer grids. It was established experimentally that the separation of the boundary layer behind the spacer grid increases the interchannel coefficient of transverse mixing of flow, and therefore, the local heat transfer. If the distance between the spacer grids is relatively small, they influence one another through the redistribution of mass flow in elementary channels. The local heat transfer behind the spacer grid varies with a distance from it. The local heat transfer can be determined accurate to within ± 3% using similar equations for any elementary cell.

Rod Bundle Heat Transfer: Steady-State Steam Cooling Experiments

2006 ◽  
Author(s):  
J. P. Spring ◽  
D. M. McLaughlin
Keyword(s):  
Heat Transfer ◽  
Heat Flux ◽  
Reynolds Number ◽  
Heat Transfer Enhancement ◽  
Local Heat Transfer ◽  
Local Heat ◽  
Transfer Enhancement ◽  
Spacer Grids ◽  
Rod Bundle ◽  
Steam Cooling

Through the joint efforts of the Pennsylvania State University and the United States Nuclear Regulatory Commission, an experimental rod bundle heat transfer (RBHT) facility was designed and built. The rod bundle consists of a 7×7 square pitch array with spacer grids and geometry similar to that found in a modern pressurized water reactor. From this facility, a series of steady-state steam cooling experiments were performed. The bundle inlet Reynolds number was varied from 1 400 to 30 000 over a pressure range from 1.36 to 4 bars (20 to 60 psia). The bundle inlet steam temperature was controlled to be at saturation for the specified pressure and the fluid exit temperature exceeded 550 °C in the highest power tests. One important quantity of interest is the local convective heat transfer coefficient defined in terms of the local bulk mean temperature of the flow, local wall temperature, and heat flux. Steam temperatures were measured at the center of selected subchannels along the length of the bundle by traversing miniaturized thermocouples. Using an analogy between momentum and energy transport, a method was developed for relating the local subchannel centerline temperature measurement to the local bulk mean temperature. Wall temperatures were measured using internal thermocouples strategically placed along the length of each rod and the local wall heat flux was obtained from an inverse conduction program. The local heat transfer coefficient was calculated from the data at each rod thermocouple location. The local heat transfer coefficients calculated for locations where the flow was fully developed were compared against several published correlations. The Weisman and El-Genk correlations were found to agree best with the RBHT steam cooling data, especially over the range of turbulent Reynolds numbers. The effect of spacer grids on the heat transfer enhancement was also determined from instrumentation placed downstream of the spacer grid locations. The local heat transfer was found to be greatest at locations immediately downstream of the grid, and as the flow moved further downstream from the grid it became more developed, thus causing the heat transfer to diminish. The amount of heat transfer enhancement was found to depend not only on the spacer grid design, but also on the local Reynolds number. It was seen that decreasing Reynolds number leads to greater heat transfer enhancement.

CFD Study on Mixing in VVER-440 Fuel Rod Bundle With Guiding Vanes in Spacer Grids

2009 ◽  
Author(s):  
Tellervo Brandt ◽  
Timo Toppila
Keyword(s):  
Turbulence Models ◽  
Heat Fluxes ◽  
Normal Operation ◽  
Navier Stokes ◽  
Pressurized Water ◽  
Spacer Grids ◽  
Fuel Rod ◽  
Rod Bundle ◽  
Computational Fluid Dynamics Cfd ◽  
Water Reactors

In this article, we study mixing in a fuel rod bundle geometry used in VVER-440 type pressurized water reactors. The flow inside the fuel assembly under normal operation is simulated using the FLUENT 6.3 computational fluid dynamics (CFD) solver. Steady state one phase Reynolds-averaged Navier-Stokes modeling is applied with two-equation turbulence models. In the first part of the article, grid independence is studied using a small submodel of the fuel rod bundle. In the second part, flow across a full-length fuel rod bundle is simulated both with constant temperature and with prescribed heat fluxes at the rod walls. Finally, in the third part of the article, we study how guiding vanes included in the spacer grids would enhance mixing in the hottest subchannels. We propose that having four guiding vanes on each side of the hexagonal spacer grid would produce most mixing in the desired locations of the fuel rod bundle.

scholarly journals CFD Turbulence Study of PWR Spacer-Grids in a Rod Bundle

2014 ◽  
Vol 2014 ◽  
pp. 1-15 ◽  
Author(s):  
C. Peña-Monferrer ◽  
J. L. Muñoz-Cobo ◽  
S. Chiva
Keyword(s):  
Mechanical Stability ◽  
Turbulence Models ◽  
Cfd Modeling ◽  
Spacer Grid ◽  
Pressurized Water ◽  
Spacer Grids ◽  
Eddy Simulation ◽  
Transient Nature ◽  
Rod Bundle ◽  
Wide Range

Nuclear fuel bundles include spacers essentially for mechanical stability and to influence the flow dynamics and heat transfer phenomena along the fuel rods. This work presents the analysis of the turbulence effects of a split-type and swirl-type spacer-grid geometries on single phase in a PWR (pressurized water reactor) rod bundle. Various computational fluid dynamics (CFD) calculations have been performed and the results validated with the experiments of the OECD/NEA-KAERI rod bundle CFD blind benchmark exercise on turbulent mixing in a rod bundle with spacers at the MATiS-H facility. Simulation of turbulent phenomena downstream of the spacer-grid presents high complexity issues; a wide range of length scales are present in the domain increasing the difficulty of defining in detail the transient nature of turbulent flow with ordinary turbulence models. This paper contains a complete description of the procedure to obtain a validated CFD model for the simulation of the spacer-grids. Calculations were performed with the commercial code ANSYS CFX using large eddy simulation (LES) turbulence model and the CFD modeling procedure validated by comparison with measurements to determine their suitability in the prediction of the turbulence phenomena.

Neutronic Analysis of Accident-Tolerant Fuel Assemblies With Silicon Carbide Cladding in Pressurized Water Reactors

2017 ◽  
Author(s):  
Zhixiong Tan ◽  
Jiejin Cai
Keyword(s):  
Silicon Carbide ◽  
Fuel Assembly ◽  
Power Distribution ◽  
Safety Margin ◽  
Severe Accident ◽  
Pressurized Water Reactors ◽  
Fuel Pellet ◽  
Pressurized Water ◽  
Neutronic Parameters ◽  
Water Reactors

After Fukushima Daiichi Nuclear Power Plant accident, alternative fuel-design to enhance tolerance for severe accident conditions becomes particularly important. Silicon carbide (SiC) cladding fuel assembly gain more safety margin as novel accident tolerant fuel. This paper focuses on the neutron properties of SiC cladding fuel assembly in pressurized water reactors. Annular fuel pellet was adopted in this paper. Two types of silicon carbide assemblies were evaluated via using lattice calculation code “dragon”. Type one was consisted of 0.057cm SiC cladding and conventional fuel. Type two was consisted of 0.089cm SiC cladding and BeO/UO2 fuel. Compared the results of SiC cladding fuel assembly neutronic parameters with conventional Zircaloy cladding fuel assembly, this paper analyzed the safety of neutronic parameters performance. Results demonstrate that assembly-level reactivity coefficient is kept negative, meanwhile, the numerical value got a relatively decrease. Other parameters are conformed to the design-limiting requirement. SiC kinds cladding show more flat power distribution. SiC cases also show the ability of reducing the enrichment of fuel pellets even though it has higher xenon concentration. These types of assembly have broadly agreement neutron performance with the conventional cladding fuel, which confirmed the acceptability of SiC cladding in the way of neutron physics analysis.

scholarly journals Numerical Investigation on the Flow Characteristics in a 17 × 17 Full-Scale Fuel Assembly

Energies ◽  
2020 ◽  
Vol 13 (2) ◽  
pp. 397 ◽  
Author(s):  
Zihao Tian ◽  
Lixin Yang ◽  
Shuang Han ◽  
Xiaofei Yuan ◽  
Hongyan Lu ◽  
...  
Keyword(s):  
Fuel Assembly ◽  
Flow Characteristics ◽  
Full Scale ◽  
Pressurized Water Reactor ◽  
Lateral Velocity ◽  
Pressurized Water ◽  
Spacer Grids ◽  
Rod Bundle ◽  
Real Size ◽  
Cfd Method

In a previous study, several computational fluid dynamics (CFD) simulations of fuel assembly thermal-hydraulic problems were presented that contained fewer fuel rods, such as 3 × 3 and 5 × 5, due to limited computer capacity. However, a typical AFA-3G fuel assembly consists of 17 × 17 rods. The pressure drop levels and flow details in the whole fuel assembly, and even in the pressurized water reactor (PWR), are not available. Hence, an appropriate CFD method for a full-scale 17 × 17 fuel assembly was the focus of this study. The spacer grids with mixing vanes, springs, and dimples were considered. The polyhedral and extruded mesh was generated using Star-CCM+ software and the total mesh number was about 200 million. The axial and lateral velocity distribution in the sub-channels was investigated. The pressure distribution downstream of different spacer grids were also obtained. As a result, an appropriate method for full-scale rod bundle simulations was obtained. The CFD analysis of thermal-hydraulic problems in a reactor coolant system can be widely conducted by using real-size fuel assembly models.

Turbulent Flow Through Spacer Grids in Rod Bundles

1998 ◽  
Vol 120 (4) ◽  
pp. 786-791 ◽  
Author(s):  
Sun Kyu Yang ◽  
Moon Ki Chung
Keyword(s):  
Turbulent Flow ◽  
Laser Doppler Velocimetry ◽  
Hydraulic Characteristics ◽  
Rod Bundles ◽  
Spacer Grid ◽  
Spacer Grids ◽  
Rod Bundle ◽  
Turbulence Decay ◽  
Skewness Factor ◽  
Flow Through

The effects of the spacer grids with mixing vanes in rod bundles on the turbulent structure were investigated experimentally. The detailed hydraulic characteristics in subchannels of a 5 × 5 rod bundle with mixing spacer grids were measured upstream and downstream of the spacer grid by using a one component LDV (Laser Doppler Velocimetry). Axial velocity and turbulent intensity, skewness factor, and flatness factor were measured. The turbulence decay behind spacer grids was obtained from measured data. The trend of turbulence decay behaves in a similar way as turbulent flow through mesh grids or screens. Pressure drop measurements were also performed to evaluate the loss coefficient for the spacer grid and the friction factor for a rod bundle.

An Estimation of the Dynamic Buckling Load for the Spacer Grid of Pressurized Water Reactor Fuel Assembly

2006 ◽  
Vol 326-328 ◽  
pp. 1603-1606 ◽  
Author(s):  
Sang Youn Jeon ◽  
Young Shin Lee
Keyword(s):  
Fuel Assembly ◽  
Dynamic Buckling ◽  
Buckling Load ◽  
Estimation Methods ◽  
Pressurized Water Reactor ◽  
Spacer Grid ◽  
Pressurized Water ◽  
Water Reactor ◽  
Spacer Grids ◽  
The Impact

This study contains an estimation of the dynamic buckling load for the spacer grid of fuel assembly in pressurized water reactor. Three different estimation methods were proposed for the calculation of the dynamic buckling loads of spacer grid. The dynamic impact tests and analyses were performed to evaluate the impact characteristics of the spacer grids and to predict the dynamic buckling load of the full size spacer grid. The estimation results were compared with the test results for the verification of the estimation methods.

scholarly journals Experimental Study of Pressure Loss in a 5 × 5–Rod Bundle With the Mixing Vane Spacer Grid

2021 ◽  
Vol 9 ◽  
Author(s):  
Wenhai Qu ◽  
Weiyi Yao ◽  
Jinbiao Xiong ◽  
Xu Cheng
Keyword(s):  
Pressure Loss ◽  
Lateral Pressure ◽  
Reynolds Numbers ◽  
Wall Friction ◽  
Geometrical Parameters ◽  
Strong Correlations ◽  
Pressurized Water Reactor ◽  
Spacer Grid ◽  
Pressurized Water ◽  
Rod Bundle

Axial and lateral pressure loss in a 5 × 5 rod–bundle with a split-type mixing vane spacer grid was experimentally measured using differential pressure transmitters at different sub-channel Reynolds numbers (Re) and orienting angles. The geometrical parameters of the 5 × 5–rod bundle are as follows: they have the same diameter (D = 9.5 mm) and pitch (p = 12.6 mm) as those of real fuel rods of a typical pressurized water reactor (PWR), with a sub-channel hydraulic diameter (Dh) of 11.78 mm. The characteristics and resistance models of pressure loss are discussed. The main axial pressure loss is caused by the spacer grid, and the spacer grid generates additional wall friction pressure loss downstream of the spacer grid. The lateral pressure loss shows strong correlations with orienting angles and distance from the spacer grid. The lateral pressure loss shows a sudden burst in the mixing vanes region and a slight augmentation at z = 3Dh. After 3Dh, the lateral pressure loss decays in an exponential way with distance from the spacer grid, and it becomes constant quickly at z = 20Dh.

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