Experimental design for the evaluation of Critical Heat Flux of small-scale pressurized nuclear reactors

2014 ◽  
pp. 959-964
Keyword(s):  
Heat Flux ◽  
Experimental Design ◽  
Critical Heat Flux ◽  
Nuclear Reactors ◽  
Small Scale

scholarly journals POST CRITICAL HEAT TRANSFER AND FUEL CLADDING OXIDATION

2016 ◽  
Vol 4 ◽  
pp. 8 ◽  
Author(s):  
Vojtěch Caha ◽  
Jakub Krejčí
Keyword(s):  
Heat Transfer ◽  
Heat Flux ◽  
Oxide Layer ◽  
Critical Heat Flux ◽  
Nuclear Reactor ◽  
Nuclear Reactors ◽  
Convection Flow ◽  
Fuel Cladding ◽  
Loss Of Coolant Accident ◽  
Nuclear Reactor Safety

The knowledge of heat transfer coefficient in the post critical heat flux region in nuclear reactor safety is very important. Although the nuclear reactors normally operate at conditions where critical heat flux (CHF) is not reached, accidents where dryout occur are possible. Most serious postulated accidents are a loss of coolant accident or reactivity initiated accident which can lead to CHF or post CHF conditions and possible disruption of core integrity. Moreover, this is also influenced by an oxide layer on the cladding surface. The paper deals with the study of mathematical models and correlations used for heat transfer calculation, especially in post dryout region, and fuel cladding oxidation kinetics of currently operated nuclear reactors. The study is focused on increasing of accuracy and reliability of safety limit calculations (e.g. DNBR or fuel cladding temperature). The paper presents coupled code which was developed for the solution of forced convection flow in heated channel and oxidation of fuel cladding. The code is capable of calculating temperature distribution in the coolant, cladding and fuel and also the thickness of an oxide layer.

scholarly journals Problems of Reliability Indicators Increase of Critical Heat Flux Calculations in the Water-Cooled Nuclear Reactors Based on the Computer Thermal-Hydraulic Codes

2018 ◽  
pp. 16-22
Author(s):  
G. Sharaevsky
Keyword(s):  
Heat Transfer ◽  
Heat Flux ◽  
Critical Heat Flux ◽  
Nuclear Reactors ◽  
Thermal Flux ◽  
Calculated Data ◽  
Two Phase ◽  
Water Steam ◽  
Heat Transfer Crisis ◽  
Computer Codes

The analysis of the current state of research and developments in the field of creation of thermal-hydraulic computer codes has been performed. The experience of creation of foreign versions of best-estimate codes was analyzed. Considerable attention is paid to the issue of critical heat flux calculation of nuclear reactors channels. It is demonstrated that now the efficiency of application of modern computer codes for estimating of the heat transfer crisis in the water-cooled nuclear reactors requires further improvement. Calculation methods for accuracy increase of predicting this thermal-hydraulic phenomenon in reactor channels are considered. The well-known methods of critical thermal flux in nuclear reactors channels have been analyzed. Peculiarities of determination of the heat transfer crisis in the forced of the vapor-water steam motion have been reviewed. Adequacy of software computer codes designed to calculate the main safety parameters of water-cooled nuclear reactors was analyzed. The idea of the physical mechanism of the heat transfer crisis under forced motion of a two-phase flow in heated channels was considered. Particular attention has been paid to analysis of experimental and calculated data on conditions of initiation of a heat transfer crisis in fuel assemblies rods.

A Brief Review of Computational Intelligence Techniques for Critical Heat Flux Prediction

2018 ◽  
Author(s):  
B. T. Jiang ◽  
Y. N. Liu
Keyword(s):  
Heat Flux ◽  
Critical Heat Flux ◽  
Computational Intelligence ◽  
Nuclear Reactor ◽  
Nuclear Reactors ◽  
Nucleate Boiling ◽  
Operating Conditions ◽  
Maximum Efficiency ◽  
Support Vector ◽  
Empirical Correlations

Critical heat flux (CHF) is one of the important design criteria of water cooled nuclear reactors and plays a key role for the safety and economics of nuclear power plants (NPPs). One of the goals of nuclear reactor design is to receive maximum efficiency under full power and its efficiency would be improved when the core exit temperature increases. From this perspective, the design of a nuclear reactor needs to take into account the appropriate thermal margin to ensure that the fuel design limits are within acceptable limits for any normal operating conditions. However, in general, CHF limits the heat flux from the fuel rods and the power capacity of the nuclear reactor. CHF refers to the transition from nucleate boiling to film boiling and causes an abrupt rise of the fuel rod surface temperature. Therefore, prediction of CHF is vital to the design and safety analysis of water cooled nuclear reactors. During the last five decades, large efforts have been carried out on the CHF prediction by many researchers. Generally, CHF prediction can be achieved in three main ways: empirical correlations, look-up tables and phenomenological models. Due to the complex nature of CHF, there is no deterministic theory for the prediction of CHF. Even the look-up tables and the empirical correlations have their own application ranges and limitations. To overcome these limitations, some computational intelligence (CI) techniques have been developed for the prediction of CHF by many researchers in the last two decades. This paper provides a brief overview of CI techniques for prediction of CHF. In this paper, the reviewed CI techniques mainly include artificial neural networks (ANNs), genetic algorithms (GAs), support vector machines (SVMs), and their hybrid models. This review also compares the strengths and weaknesses of several CI techniques and provides basic technical support for future selection of appropriate methods by those involved in the field.

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