scholarly journals Rod bundle test facility instrumentation description, calibration, and uncertainty analysis, PWR blowdown heat transfer project. Final report

1977 ◽  
Author(s):  
A. H. Hsia ◽  
F. R. Hubbard, III ◽  
R. E. Schneider
Keyword(s):  
Heat Transfer ◽  
Uncertainty Analysis ◽  
Test Facility ◽  
Final Report ◽  
Rod Bundle

Validation and Uncertainty Analysis of the Thermal Hydraulics Module of SOCRAT-BN Code on the Rod Bundle Experiment

2015 ◽  
Vol 1084 ◽  
pp. 717-721
Author(s):  
Yulia Vinogradova ◽  
Nikolai Ryzhov ◽  
Ruslan Chalyy
Keyword(s):  
Heat Transfer ◽  
Uncertainty Analysis ◽  
Reactor Core ◽  
Thermal Hydraulics ◽  
Fast Reactors ◽  
Model Uncertainties ◽  
Fuel Rod ◽  
Rod Bundle ◽  
Code Model ◽  
Design Basis

SOCRAT-BN code is developed for the analysis of design and beyond design basis accidents at sodium cooled fast reactors. To simulate the behavior of the coolant in the reactor core heat transfer and friction in rod bundle geometry are required to consider. The article describes the validation of the code SOCRAT-BN on the experiment with fuel rod imitators in the triangular geometry with wire-wound taking into account experiment and some code model uncertainties.

A Research Study on Internal Corrosion of High-Pressure Boilers

1968 ◽  
Vol 90 (1) ◽  
pp. 21-37 ◽  
Author(s):  
P. Goldstein
Keyword(s):  
Heat Transfer ◽  
High Pressure ◽  
Water Treatment ◽  
Research Study ◽  
Test Facility ◽  
Final Report ◽  
Boiler Water ◽  
Internal Corrosion ◽  
The Third ◽  
Simulated Seawater

The following report is the third and last in a series describing the progress of “A Research Study on Internal Corrosion of High Pressure Boilers.” The first report described the background, scope, and organization of the program as well as the test facility. The second report discussed the methods of testing and the results of the first six runs. This final report describes the results of the last six tests and discusses the conclusions drawn from all of Phases II and III. The scope and an outline of seven tests composing the newly scheduled Phase IV program are also included. The results of runs with three types of boiler water treatment, fouled heat transfer surfaces, and conditions simulating fresh water and seawater condenser leakage are included. Data relating to deposition and corrosion in these environments are presented with particular emphasis on the severe corrosion experienced with simulated seawater condenser leakage.

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.

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.

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