scholarly journals Rod bundle test facility description for PWR blowdown heat transfer project. Technical report 2

1975 ◽  
Author(s):  
H. Muenchow
Keyword(s):  
Heat Transfer ◽  
Test Facility ◽  
Rod Bundle ◽  
Technical Report

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.

Experimental Verification of the Performance of a Droplet Injection System for Use in a Rod Bundle Test Facility

2001 ◽  
Author(s):  
A. J. Ireland ◽  
E. Rosal ◽  
L. E. Hochreiter ◽  
F. B. Cheung
Keyword(s):  
Heat Transfer ◽  
Nuclear Reactor ◽  
Injection System ◽  
Test Facility ◽  
Film Boiling ◽  
Size Distributions ◽  
Dispersed Flow ◽  
Rod Bundle ◽  
Drop Size Distributions ◽  
Flow Film Boiling

Abstract A droplet injection system has been developed for use in a rod bundle heat transfer test facility designed specifically for the study of dispersed flow film boiling during reflood transients in a nuclear reactor. Three different injectors having various pitch configurations and hole patterns were tested. The drop field produced by each was characterized using a laser-assisted size measurement technique. Appropriate mean diameter and drop size distributions that closely simulate the drop field encountered in a reactor bundle assembly under reflood conditions was obtained. The drop injection system developed can readily be employed in the rod bundle test facility to investigate the droplet heat transfer in the dispersed flow film boiling regime.

A Blind, Numerical Benchmark Study on Supercritical Water Heat Transfer Experiments in a 7-Rod Bundle

2016 ◽  
Vol 2 (2) ◽  
Author(s):  
M. Rohde ◽  
J. W. R. Peeters ◽  
A. Pucciarelli ◽  
A. Kiss ◽  
Y. F. Rao ◽  
...  
Keyword(s):  
Heat Transfer ◽  
Supercritical Water ◽  
Real Life ◽  
Predictive Ability ◽  
Test Facility ◽  
Test Case ◽  
Significant Deterioration ◽  
Rod Bundle ◽  
Sst Turbulence Model ◽  
Water Reactors

Heat transfer in supercritical water reactors (SCWRs) shows a complex behavior, especially when the temperatures of the water are near the pseudocritical value. For example, a significant deterioration of heat transfer may occur, resulting in unacceptably high cladding temperatures. The underlying physics and thermodynamics behind this behavior are not well understood yet. To assist the worldwide development in SCWRs, it is therefore of paramount importance to assess the limits and capabilities of currently available models, despite the fact that most of these models were not meant to describe supercritical heat transfer (SCHT). For this reason, the Gen-IV International Forum initiated the present blind, numerical benchmark, primarily aiming to show the predictive ability of currently available models when applied to a real-life application with flow conditions that resemble those of an SCWR. This paper describes the outcomes of ten independent numerical investigations and their comparison with wall temperatures measured at different positions in a 7-rod bundle with spacer grids in a supercritical water test facility at JAEA. The wall temperatures were not known beforehand to guarantee the blindness of the study. A number of models have been used, ranging from a one-dimensional (1-D) analytical approach with heat transfer correlations to a RANS simulation with the SST turbulence model on a mesh consisting of 62 million cells. None of the numerical simulations accurately predicted the wall temperature for the test case in which deterioration of heat transfer occurred. Furthermore, the predictive capabilities of the subchannel analysis were found to be comparable to those of more laborious approaches. It has been concluded that predictions of SCHT in rod bundles with the help of currently available numerical tools and models should be treated with caution.

Sign in / Sign up

Export Citation Format

Share Document