CRITICAL HEAT FLUX ON A CYLINDER OF LARGE DIAMETER IN A CROSS FLOW

1986 ◽  
Author(s):  
G. Meyer ◽  
E. S. Gaddis ◽  
A. Vogelpohl
Keyword(s):  
Heat Flux ◽  
Critical Heat Flux ◽  
Large Diameter ◽  
Cross Flow

Experimental Evaluation of Critical Heat Flux in Downward-Facing Boiling on SS304 L Flat Plate Relevant to In-Calandria Retention in PHWRs

2020 ◽  
Vol 6 (3) ◽  
Author(s):  
Sumit V. Prasad ◽  
A. K. Nayak
Keyword(s):  
Heat Transfer ◽  
Heat Flux ◽  
Natural Convection ◽  
Flat Plate ◽  
Critical Heat Flux ◽  
Outer Surface ◽  
Flat Surface ◽  
Large Diameter ◽  
Convection Heat Transfer ◽  
Heat Removal

Abstract Retention of corium inside the calandria vessel (CV) by externally cooling it by calandria vault water is essential to mitigate severe accidents in pressurized heavy water reactor (PHWR). The thermal failure of CV can be prevented by effective decay heat removal on the outer surface of CV using vault water, which depends on the heat transfer behavior from the outer surface of CV to the vault water. Determination of limiting heat removal capability of vault subcooled water through outer surface of CV is very important. Since the CV has a very large diameter and length, the bottom most part of the CV almost behaves as a flat plate with downward natural convection boiling heat transfer. The natural convection heat transfer is lesser on the flat surface as compared to the curved surface of the CV. Thus, the critical heat flux (CHF) on the flat surface under downward boiling condition is the limiting CHF of the CV under external surface boiling scenario. In order to estimate CHF in this configuration with local boiling, experiments were carried out on a downward facing SS304 L flat plate simulating the conditions of CV of 700 MWel Indian PHWR. The pool boiling CHF obtained in this study is also compared with other earlier works.

scholarly journals Application of the Critical Heat Flux Look-Up Table to Large Diameter Tubes

2013 ◽  
Vol 2013 ◽  
pp. 1-10 ◽  
Author(s):  
M. El Nakla ◽  
M. Habib ◽  
W. Ahmed ◽  
A. Al-Sarkhi ◽  
R. Ben Mansour ◽  
...  
Keyword(s):  
Heat Flux ◽  
Critical Heat Flux ◽  
Correction Factor ◽  
Large Diameter ◽  
Horizontal Tube ◽  
High Heat ◽  
Vertical Tube ◽  
High Heat Flux ◽  
Look Up Table ◽  
Boiler Tubes

The critical heat flux look-up table was applied to a large diameter tube, namely 67 mm inside diameter tube, to predict the occurrence of the phenomenon for both vertical and horizontal uniformly heated tubes. Water was considered as coolant. For the vertical tube, a diameter correction factor was directly applied to the 1995 critical heat flux look-up table. To predict the occurrence of critical heat flux in horizontal tube, an extra correction factor to account for flow stratification was applied. Both derived tables were used to predict the effect of high heat flux and tube blockage on critical heat flux occurrence in boiler tubes. Moreover, the horizontal tube look-up table was used to predict the safety limits of the operation of boiler for 50% allowable heat flux.

Study on the Prediction of DNB-Type Critical Heat Flux in Rod Bundle Under Motion Conditions

2018 ◽  
Author(s):  
Siyang Huang ◽  
Xiaoyan Wang ◽  
Wenxi Tian ◽  
Ronghua Chen ◽  
Junmei Wu ◽  
...  
Keyword(s):  
Heat Flux ◽  
Critical Heat Flux ◽  
Nuclear Reactor ◽  
Cross Flow ◽  
Nucleate Boiling ◽  
Heated Wall ◽  
Coupled Analysis ◽  
Flow Conditions ◽  
Local Flow ◽  
Rod Bundle

In the nuclear reactor design, the critical heat flux (CHF) is one of the most important parameters for the reactor safety analysis. The occurrence of CHF will cause a sharp increase in the fuel rod surface temperature, which will result in the failure of fuel claddings and damage of the core. The CHF depends on the local flow conditions and the geometry of the flow channels, which makes the prediction of CHF in a fuel assembly more difficult when considering the cross flow between neighboring channels, spacer grids and mixing vanes. In this paper, the departure from nucleate boiling (DNB) type CHF in rod bundles under motion conditions is investigated based on the coupled analysis of the subchannel method and a CHF mechanism model, namely the liquid sublayer dryout model. The liquid sublayer dryout model assumes that there is a thin liquid sublayer underneath a vapor blanket formed by the coalescence of small bubbles near the heated wall. The dryout of this sublayer is considered as the CHF occurrence. In the liquid sublayer dryout model, sublayer thickness, velocity and length of the vapor blanket are three crucial parameters. In present research, the subchannel code calculates the local flow conditions for the rod bundle and provides input parameters for the liquid sublayer dryout model to predict CHF. In order to verify the method above, the predicted results are compared with the CHF Look-Up Table 2006 (LUT-2006) and a reasonable agreement can be achieved. In addition, the effects of rod bundle inlet coolant mass flow rate, subcooling and motion conditions on the CHF are analyzed.

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