Droplet Measurement in a Rod Bundle Geometry for a Reflood Heat Transfer Test

2010 ◽  
Author(s):  
H. K. Cho ◽  
K. Y. Choi ◽  
S. Cho ◽  
C.-H. Song
Keyword(s):  
Heat Transfer ◽  
Droplet Size ◽  
Superheated Steam ◽  
Image Processing Technique ◽  
Test Facility ◽  
Flow Conditions ◽  
Spacer Grid ◽  
Rod Bundle ◽  
Droplet Behavior ◽  
Break Up

During the reflood phase of a postulated loss of coolant accident in a nuclear reactor, the entrainment of liquid droplets can occur at a quench front of reflooding water. It is widely recognized that the behavior of the entrained droplet crucially affects the reflood heat transfer phenomena by decreasing the superheated steam temperature and interacting with a rod bundle and spacer grids. For this reason, various experimental and numerical studies have been performed to examine droplet behavior such as the droplet size, velocity and droplet fraction inside a rod array. In this study, an experiment on the droplet behavior inside a heated rod bundle has been performed. The experiment was focused on the break-up of droplets induced by a spacer grid in a rod bundle geometry, which results in the increase of the interfacial heat transfer between droplets and superheated steam. A 6×6 rod bundle test facility in Korea Atomic Energy Research Institute was used for the experiment. Steam was supplied by an external boiler into the bottom of the test channel, and a droplet injection nozzle was equipped instead of simulating a quench front of reflooding water. The major measuring parameters of the experiment were the droplet size and velocity, and these were measured by a high-speed camera and a digital image processing technique. A series of experiments were conducted with various flow conditions of a steam injection velocity, heater temperature, droplet size and droplet flow rate. The experiments provided the data on the change of the Sauter mean diameter of droplets after collision with a spacer grid depending on flow conditions. Moreover, the data was analyzed with a droplet break-up model by a spacer grid which was implemented into a thermal hydraulic analysis code, COBRA-TF.

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.

Preliminary Investigation on Convective Heat Transfer of Rod Bundle Flow With Spacer Grid for Nuclear Fuel Assembly Application

2015 ◽  
Author(s):  
Chi Young Lee ◽  
Chang Hwan Shin ◽  
Wang Kee In ◽  
Dong Seok Oh ◽  
Tae Hyun Chun
Keyword(s):  
Heat Transfer ◽  
Fuel Assembly ◽  
Convective Heat Transfer ◽  
Convective Heat ◽  
Heat Transfer Performance ◽  
Axial Flow ◽  
Transfer Performance ◽  
Spacer Grid ◽  
Rod Bundle ◽  
Nuclear Fuel Assembly

The convective heat transfer of rod bundle flow with spacer grid was investigated preliminarily for nuclear reactor core application. As the test fluid, the water was used. To simulate the nuclear fuel assembly, 4×4 rod bundle with P/D (=pitch between rods/rod diameter) of ∼1.35 was prepared together with a spacer grid with twist-mixing vane. A single heated section with five thermocouples embedded in the surface along the circumferential direction was installed around the center subchannel. The measurements of wall temperatures were carried out upstream and downstream of spacer grid. For the rod bundle flow at the inlet of spacer grid (i.e., upstream of spacer grid), the wall temperatures at the gap and subchannel centers exhibited the higher and lower, respectively, which was because in the subchannel center, the axial flow velocity became higher, as compared with the gap center. On the other hand, downstream of spacer grid, the rod wall toward the tip of twist-mixing vane showed the lowest temperature in the measurements along the circumferential direction of rod wall. Near the twist-mixing vane, the averaged wall temperature was observed to be remarkably low, which implies that the twist-mixing vane is an effective tool to enhance the convective heat transfer performance. However, along the axial flow direction behind the spacer grid, the averaged wall temperatures became to increase, and the enhancement of convective heat transfer performance by mixing vane faded away.

scholarly journals Effect of Flow Blockage on the Coolability during Reflood in a 2×2 Rod Bundle

2014 ◽  
Vol 2014 ◽  
pp. 1-9 ◽  
Author(s):  
Kihwan Kim ◽  
Byung-Jae Kim ◽  
Young-Jung Youn ◽  
Hae-Seob Choi ◽  
Sang-Ki Moon ◽  
...  
Keyword(s):  
Heat Transfer ◽  
Test Facility ◽  
Pressurized Water Reactor ◽  
Two Phase ◽  
Flow Area ◽  
Pressurized Water ◽  
Fuel Rods ◽  
Rod Bundle ◽  
Loss Of Coolant Accident ◽  
Two Phase Heat Transfer

During the reflood phase of a large-break loss-of-coolant accident (LBLOCA) in a pressurized-water reactor (PWR), the fuel rods can be ballooned or rearranged owing to an increase in the temperature and internal pressure of the fuel rods. In this study, an experimental study was performed to understand the thermal behavior and effect of the ballooned region on the coolability using a 2 × 2 rod bundle test facility. The electrically heated rod bundle was used and the ballooning shape of the rods was simulated by superimposing hollow sleeves, which have a 90% blockage ratio. Forced reflood tests were performed to examine the transient two-phase heat transfer behavior for different reflood rates and rod powers. The droplet behaviors were also investigated by measuring the velocity and size of droplets near the blockage region. The results showed that the heat transfer was enhanced in the downstream of the blockage region, owing to the reduced flow area of the subchannel, intensification of turbulence, and deposition of the droplet.

Computational Fluid Dynamics Simulation of Flow-Mixing and Heat Transfer in 4 × 4 Rod Bundle With a Twist-Vane Spacer Grid

2019 ◽  
Vol 5 (4) ◽  
Author(s):  
Ganesh Lal Kumawat ◽  
Anuj Kumar Kansal ◽  
Naresh Kumar Maheshwari ◽  
Avaneesh Sharma
Keyword(s):  
Heat Transfer ◽  
Fluid Dynamics ◽  
Computational Fluid Dynamics ◽  
Atomic Energy ◽  
Cfd Simulations ◽  
Spacer Grid ◽  
Fuel Rod ◽  
Rod Bundle ◽  
Flow Mixing ◽  
Computational Fluid Dynamics Cfd

The clearance between fuel rods is maintained by spacer grid or helical wire wrap. Thermal-hydraulic characteristics inside fuel rod bundle are strongly influenced by the spacer grid geometry and the bundle pitch-to-diameter (P/D) ratio. This includes the maximum fuel temperature, critical heat flux, as well as pressure drop through the fuel bundle. An understanding of the detailed structure of flow mixing and heat transfer in a fuel rod bundle geometry is therefore an important aspect of reactor core design, both in terms of the reactor's safe and reliable operation, and with regard to optimum power extraction. In this study, computational fluid dynamics (CFD) simulations are performed to investigate isothermal turbulent flow mixing and heat transfer behavior in 4 × 4 rod bundle with twist-vane spacer grid with P/D ratio of 1.35. This work is carried out under International Atomic Energy Agency (IAEA) co-ordinated research project titled as “Application of Computational Fluid Dynamics (CFD) Codes for Nuclear Power Plant Design.” CFD simulations are performed using open source CFD code OpenFOAM. Numerical results are compared with experimental data from Korea Atomic Energy Research Institute (KAERI) and found to be in good agreement.

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