Forced-Convective Post-CHF Heat Transfer and Quenching

1982 ◽  
Vol 104 (1) ◽  
pp. 48-54 ◽  
Author(s):  
R. A. Nelson
Keyword(s):  
Heat Transfer ◽  
Heat Flux ◽  
Homogeneous Nucleation ◽  
Nuclear Reactor ◽  
Reactor Safety ◽  
Fuel Rod ◽  
Different Types ◽  
Nuclear Reactor Safety ◽  
Transfer Problems ◽  
Prediction Capability

Mechanisms in the postcritical heat flux region that provide understanding and qualitative prediction capability for several current force-convective heat-transfer problems are discussed. In the area of nuclear reactor safety, the mechanisms are important in the prediction of fuel rod cooldown and quenches for the reflood phase, blowdown phase, and possibly some operational transients with dryout. Results using the mechanisms to investigate forced-convective quenching are presented. Data reduction of quenching experiments is discussed, and the way in which the quenching transient may affect the results of different types of quenching experiments is investigated. This investigation provides an explanation of how minimum wall superheats greater than the homogeneous nucleation temperature result, as well as how these may be either hydrodynamically or thermodynamically controlled.

Heat Transfer Problems in Nuclear Reactor Safety

1976 ◽  
Vol 190 (1) ◽  
pp. 215-224
Author(s):  
W. B. Hall
Keyword(s):  
Heat Transfer ◽  
Safety Assessment ◽  
Nuclear Reactor ◽  
Reactor Safety ◽  
Operating Characteristics ◽  
Nuclear Reactor Safety ◽  
Thermal Explosions ◽  
Transient Boiling ◽  
Transfer Problems

Reactor safety assessment is a highly specialized topic which, in many of its aspects, depends heavily on a satisfactory understanding of a wide variety of heat transfer phenomena. It is the aim of the paper to air some of these problems outside the ranks of the reactor safety specialists. Typical liquid-cooled reactors, their operating characteristics, and some heat transfer aspects of their safety assessment are discussed: for example, transient boiling, quenching of hot surfaces and thermal explosions.

Calculation of Critical Heat Flux Using an Inverse Heat Transfer Method to Support TREAT Experiment Analysis

2020 ◽  
Author(s):  
Robert Armstrong ◽  
Charles Folsom ◽  
Connie Hill ◽  
Colby Jensen
Keyword(s):  
Heat Transfer ◽  
Heat Flux ◽  
Critical Heat Flux ◽  
Nuclear Reactor ◽  
National Laboratory ◽  
Test Facility ◽  
Inverse Heat Transfer ◽  
Transfer Method ◽  
Nuclear Reactor Safety ◽  
Accident Scenarios

Abstract Heat transfer between cladding and coolant during transient scenarios remains a critical area of uncertainty in understanding nuclear reactor safety. To advance the understanding of transient and accident scenarios involving critical heat flux (CHF), an in-pile experiment for the Transient Reactor Test facility (TREAT) at Idaho National Laboratory (INL) was developed. The experiment, named CHF-Static Environment Rodlet Transient Test Apparatus (CHF-SERTTA), consists of a hollow borated stainless-steel heater rod submerged in a static water pool heated via the (n, α) reaction in boron-10. This paper presents a novel inverse heat transfer method to determine CHF by using the optimization and uncertainty software Dakota to calibrate a RELAP5-3D model of CHF-SERTTA to temperature measurements obtained from a thermocouple welded to the surface of the rod.

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.

Nuclear Reactor Safety Heat Transfer

1982 ◽  
Vol 58 (1) ◽  
pp. 122-123
Author(s):  
Clifford J. Cremers
Keyword(s):  
Heat Transfer ◽  
Nuclear Reactor ◽  
Reactor Safety ◽  
Nuclear Reactor Safety
Sign in / Sign up

Export Citation Format

Share Document