scholarly journals Numerical and experimental investigation of the water flow through PWR spacer grids

2021 ◽  
Vol 8 (3A) ◽  
Author(s):  
Higor Fabiano Pereira de Castro ◽  
Guilherme Augusto Moura Vidal ◽  
Tiago Augusto Santiago Vieira ◽  
Vitor Vasconcelos Araújo Silva ◽  
Daniel De Almeida Magalhães Campolina ◽  
...  
Keyword(s):  
Water Flow ◽  
Technology Development ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Spacer Grids ◽  
Rod Bundle ◽  
Ansys Cfx ◽  
Flow Through ◽  
Belo Horizonte ◽  
Main Components

Spacer grids are one of main components of a Pressurized Water Reactor (PWR) fuel assembly. They are able to improve heat transfer from rod bundles to the water flow by increasing turbulence and mixture of this flow. On the other hand the pressure drop increases because spacer grids. Experimental and Computational Fluid Dynamics (CFD) analysis have been used to understand how spacer grids affect the water flow. This analysis is important to improve spacer grids thermal-hydraulic performance. This paper aims to investigate numerically and experimentally the water flow through PWR spacer grids. The numerical and experimental procedures have been developed for a 5x5 rod bundle with spacer grids at the Nuclear Technology Development Center (CDTN) in Belo Horizonte, Brazil. At CDTN, measurements of the velocity components are acquired with a 2D LDV (Laser Doppler Velocimetry) system and the numerical results are obtained using ANSYS CFX code. The measurements are obtained at one height downstream from a spacer grid and compared to CFD simulations for a flow rate at Reynolds number of 5.4x104 . Results show good agreement between both methodologies. The great repeatability and low experimental uncertainty evaluated (< 1.24%) in this work can be used to validate other CFD codes.

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.

Leaching Studies of Cement-Glass Package Containing Sodium Borate

1986 ◽  
Vol 84 ◽  
Author(s):  
Masahiro Okamoto ◽  
Koichi Chino ◽  
Tsutomu Baba ◽  
Tatsuo Izumida ◽  
Fumio Kawamura ◽  
...  
Keyword(s):  
Sodium Silicate ◽  
Liquid Waste ◽  
Sodium Borate ◽  
Pressurized Water Reactor ◽  
X Ray Diffraction ◽  
Adsorption Ability ◽  
Pressurized Water ◽  
X Ray ◽  
Solidification Reaction ◽  
Main Components

AbstractA new solidification technique using cement-glass, which is a mixture of sodium silicate, cement, additives, and initiator of the solidification reaction, was developed for sodium borate liquid waste generated from pressurized water reactor (PWR) plants. The cement-glass could solidify eight times as much sodium borate as cement could, because the solidifying reaction of the cement-glass is not hindered by borate ions.The reaction mechanism of sodium silicate and phosphoric silicate (initiator), the main components of cement-glass, was studied through X-ray diffraction and compressive strength measurements. It was found that three- dimensionally bonded silicon dioxide was produced by polymerization of the two silicates. The leaching ratio of cesium from the cement-glass package was one-tenth that of the cement one. This low value was attributed to a high cesium adsorption ability of the cement-glass and it could be theoretically predicted accordingly.

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.

Numerical Prediction of Void Distribution in Two-Phase Turbulent Flow in a Sub-Channel

2009 ◽  
Author(s):  
Tsutomu Ikeno ◽  
Tatsuya Sasakawa ◽  
Isao Kataoka
Keyword(s):  
Turbulent Flow ◽  
Transport Equation ◽  
Nucleate Boiling ◽  
Pressurized Water Reactor ◽  
Two Phase ◽  
Simulation Code ◽  
Pressurized Water ◽  
Void Fractions ◽  
Rod Bundle ◽  
Void Distribution

Numerical simulation code for predicting void distribution in two-phase turbulent flow in a sub-channel was developed. The purpose is to obtain a profile of void distribution in the sub-channel. The result will be used for predicting a heat flux at departure from nucleate boiling (DNB) in a rod bundle for the pressurized water reactor (PWR). The fundamental equations were represented by a generalized transport equation, and the transport equation was transformed onto the generalized coordinate system fitted to the rod surface and the symmetric lines in the sub-channel. Using the finite-volume method the transport equation was discretized for the SIMPLE algorism. The flow field and void fraction at the steady state were calculated for different average void fractions. The computational result for atmospheric pressure condition was successfully compared with experimental data. Sensitivity analysis for the PWR condition was performed, and the result showed that the secondary flow slightly contributed to homogenizing the void distribution.

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.

Experimental and Numerical Modeling of Transition Matrix From Momentum to Buoyancy-Driven Flow in a Pressurized Water Reactor

2008 ◽  
Vol 131 (1) ◽  
Author(s):  
Thomas Höhne ◽  
Sören Kliem ◽  
Roman Vaibar
Keyword(s):  
Mass Flow ◽  
Transition Matrix ◽  
Turbulence Models ◽  
Wire Mesh ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Flow Rates ◽  
Water Reactor ◽  
Ansys Cfx ◽  
Mass Flow Rates

The influence of density differences on the mixing of the primary loop inventory and the emergency core cooling (ECC) water in the cold leg and downcomer of a pressurized water reactor (PWR) was analyzed at the Rossendorf coolant mixing (ROCOM) test facility. This paper presents a matrix of ROCOM experiments in which water with the same or higher density was injected into a cold leg of the reactor model with already established natural circulation conditions at different low mass flow rates. Wire-mesh sensors measuring the concentration of a tracer in the injected water were installed in the cold leg, upper and lower part of the downcomer. A transition matrix from momentum to buoyancy-driven flow experiments was selected for validation of the computational fluid dynamics software ANSYS CFX. A hybrid mesh with 4×106 elements was used for the calculations. The turbulence models usually applied in such cases assume that turbulence is isotropic, whilst buoyancy actually induces anisotropy. Thus, in this paper, higher order turbulence models have been developed and implemented, which take into account that anisotropy. Buoyancy generated source and dissipation terms were proposed and introduced into the balance equations for the turbulent kinetic energy. The results of the experiments and of the numerical calculations show that mixing strongly depends on buoyancy effects: At higher mass flow rates (close to nominal conditions) the injected slug propagates in the circumferential direction around the core barrel. Buoyancy effects reduce this circumferential propagation with lower mass flow rates and/or higher density differences. The ECC water falls in an almost vertical path and reaches the lower downcomer sensor directly below the inlet nozzle. Therefore, density effects play an important role during natural convection with the ECC injection in PWR and should be also considered in pressurized thermal shock scenarios. ANSYS CFX was able to predict the observed flow patterns and mixing phenomena quite well.

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.

Evaluation of Entrance Blockage of Inner Channel in Dual-Cooled Annular Nuclear Fuel

2011 ◽  
Author(s):  
C. H. Shin ◽  
T. H. Chun ◽  
D. S. Oh ◽  
W. K. In
Keyword(s):  
Side Surface ◽  
Nucleate Boiling ◽  
Pressurized Water Reactor ◽  
Bypass Flow ◽  
Pressurized Water ◽  
Channel Blockage ◽  
Key Issues ◽  
Flow Through ◽  
Side Flow ◽  
Inner Channel

The Korea Atomic Energy Research Institute (KAERI) has been developing a dual-cooled annular fuel for the power uprate of 20% in an optimized pressurized water reactor (PWR) in Korea, OPR1000. The dual-cooled annular fuel is configured to allow the coolant flow through the inner channel as well as the outer channel. Several thermal-hydraulic issues exist for the application of the dual-cooled annular fuel to OPR1000. One of the key issues is the hypothetical event of inner channel blockage because the inner channel is an isolated flow channel without the coolant mixing between the neighboring flow channels. The inner channel blockage could cause the Departure from Nucleate Boiling (DNB) in the inner channel that eventually results in a fuel failure. A long lower end cap for the annular fuel was invented to provide flow holes by perforating the side surface of the end cap body. The side holes in the lower end cap are expected to supply a minimum coolant in the inner channel in order to prevent the DNB occurrence in the event of partial or even complete blockage of the inner channel entrance. But due to very unusual shape of the lower end cap, it is difficult to estimate the flow resistance of the side flow holes using empirical equations available in the open literatures. Experimental and computational fluid dynamics (CFD) study were performed to investigate the bypass flow through the side holes of the end cap in the case of complete entrance blockage of the inner channel. The form loss coefficient in the side holes was also estimated by using the pressure drop along the bypass flow path.

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