AREVA’s Test Facility KATHY: Robust Critical Heat Flux Measurements, a Prerequisite for Reliable CHF Prediction

2014 ◽  
Author(s):  
O. Wieckhorst ◽  
J. Kronenberg ◽  
H. Gabriel ◽  
S. Opel ◽  
D. Kreuter ◽  
...  
Keyword(s):  
Heat Flux ◽  
Fuel Assembly ◽  
Critical Heat Flux ◽  
Development Process ◽  
Heat Removal ◽  
Test Facility ◽  
Hydraulic Facility ◽  
Key Factor ◽  
Unique Experience ◽  
High Level

The primary tool for assuring the heat removal from the fuel design’s rod surfaces is properly represented in the numerical simulations of a LWR fuel assembly design is the critical heat flux (CHF) or dryout correlation. During the last decade, AREVA has compiled unique experience in correlation development that has led to an improved development process to meet increased technical challenges. This is based upon the high level of expertise in CHF measurements for PWR and BWR fuel assembly designs gained by AREVA at its KATHY facility (KArlstein Thermal HYdraulic facility). The utilization of KATHY in conjunction with this improved development process is a key factor in ensuring reliable CHF prediction for safety analysis application. This paper describes the capabilities of the KATHY loop and the process used by AREVA to attain high quality CHF measurements.

Development of an Experimental Facility for Investigating Critical Heat Flux of Saturated Flow Boiling of Refrigerant-123 in Microchannels

2005 ◽  
Author(s):  
Wai Keat Kuan ◽  
Satish G. Kandlikar
Keyword(s):  
Heat Flux ◽  
Critical Heat Flux ◽  
Flow Boiling ◽  
Test Facility ◽  
Uniform Heat Flux ◽  
Inlet Temperature ◽  
Experimental Facility ◽  
Copper Block ◽  
Saturated Flow ◽  
Cartridge Heater

An experimental facility is developed to investigate critical heat flux (CHF) of saturated flow boiling of Refrigerant-123 (R-123) in microchannels. Six parallel Microchannels with cross sectional area of 0.2 mm × 0.2 mm are fabricated on a copper block, and a Polyvinyl Chloride (PVC) cover is then placed on top of the copper block to serve as a transparent cover through which flow patterns and boiling phenomena could be observed. A resistive cartridge heater is used to provide a uniform heat flux to the microchannels. The experimental test facility is designed to accommodate test sections with different microchannel geometries. The mass flow rate, inlet pressure, inlet temperature of Refrigerant-123, and the electric current supplied to the resistive cartridge heater are controlled to provide quantitative information near the CHF condition in microchannels. A high-speed camera is used to observe and interpret flow characteristics of CHF condition in microchannels.

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.

Numerical and Parametric Investigation of the Effect of Heat Spreading On Boiling of a Dielectric Liquid for Immersion Cooling of Electronics

2021 ◽  
Author(s):  
Wei Tong ◽  
Alireza Ganjali ◽  
Omidreza Ghaffari ◽  
Chady Alsayed ◽  
Luc Frechette ◽  
...  
Keyword(s):  
Heat Transfer ◽  
Heat Flux ◽  
Heat Transfer Coefficient ◽  
Surface Temperature ◽  
Transfer Coefficient ◽  
Critical Heat Flux ◽  
Heat Removal ◽  
Nucleate Boiling ◽  
Input Power ◽  
Immersion Cooling

Abstract In a two-phase immersion cooling system, boiling on the spreader surface has been experimentally found to be non-uniform, and it is highly related to the surface temperature and the heat transfer coefficient. An experimentally obtained temperature-dependent boiling heat transfer coefficient has been applied to a numerical model to investigate the spreader's cooling performance. It is found that the surface temperature distribution becomes less uniform with higher input power. But it is more uniform when the thickness is increased. By defining the characteristic temperatures that represent different boiling regimes on the surface, the fraction of the surface area that has reached the critical heat flux has been numerically calculated, showing that increasing the thickness from 1 mm to 6 mm decreases the critical heat flux reached area by 23% at saturation liquid temperatures. Therefore, on the thicker spreader, more of the surface is utilized for nucleate boiling while localized hot regions that lead to surface dry-out are avoided. At a base temperature of 90 oC, the optimal thickness is found to be 4 mm, beyond which no significant improvement in heat removal can be obtained. Lower coolant temperatures can further increase the heat removal; it is reduced from an 18% improvement in the input power for the 1 mm case to only 3% in the 6 mm case for a coolant temperature drop of 24 oC. Therefore, a trade-off exists between the cost of maintaining the low liquid temperature and the increased heat removal capacity.

Experimental Investigation of Boil-Off-Scenario in BWR Spent Fuel Pools

2013 ◽  
Author(s):  
Steffen Schulz ◽  
Christoph Schuster ◽  
Antonio Hurtado
Keyword(s):  
Fuel Assembly ◽  
Great Influence ◽  
Heat Removal ◽  
Heat Fluxes ◽  
Test Facility ◽  
Spent Fuel ◽  
Radial Temperature ◽  
Analytical Approximations ◽  
Radial Heat ◽  
Inner Channel

Accident scenarios in spent fuel pools (SFP) were not a main topic in nuclear safety studies so far. Heat fluxes are low and the timeframe for counteractions in case of a loss of cooling is long. However, because of their huge activity inventory safety of spent fuel pools should be analyzed with great intensity. Beside some analytical approximations from the U.S. Nuclear Regulatory Commission (NRC) there is only few experimental data about loss of coolant accidents in a SFP [1, 3, 4, 5, 6]. After the reactor accident in Fukushima Daiichi NPP this issue became omnipresent in public perception. The presented paper investigates boil-off scenario of a boiling water reactor (BWR) SFP after a station blackout with focus on experimental findings. Since 2006 experiments for SFP boil-off scenarios were conducted at the Technische Universität Dresden [7]. For a detailed investigation under realistic geometrical fuel assembly conditions the test facility ADELA-II was built in 2011. It consists of two channels. The inner channel simulates a quarter of a BWR fuel element with a 24 rod bundle. In the outer channel 8 more heating rods were assembled to reduce heat losses from the inner channel and to simulate the surrounding. With a heated length of 3760 mm and a width of 120 mm it was designed in an axial and radial scale of 1:1. The heating profile is based on the heat flux profile of a fuel rod with low burn-up. To measure the detailed axial and radial temperature profiles 113 thermocouples are mounted. With these data conclusions for heat and mass transfer can be made with special regard to convection inside the assembly. Boil-off experiments at the ADELA-II facility with a supplied heat of 20 Watts per rod lead to rod surface peak temperatures of about 479 °C. Due to the high temperature and narrowed geometry the axial heat transport is harmed and radial heat conduction had a great influence on the test results. Accordingly, quasi adiabatic boil-off could not be verified. Thus, coolability of a fuel assembly strongly depends on the (radial) heat removal from the fuel assembly box to the surrounding. Further investigations have to be done for a better consideration of neighbored fuel elements and global thermohydraulic effects in the spent fuel pool.

Microporous Coatings and Enhanced Critical Heat Flux for Downward Facing Boiling During Passive Emergency Reactor Cooling

2017 ◽  
Vol 139 (5) ◽  
Author(s):  
Albert E. Segall ◽  
Faruk A. Sohag ◽  
Faith R. Beck ◽  
Lokanath Mohanta ◽  
Fan-Bill Cheung ◽  
...  
Keyword(s):  
Heat Flux ◽  
Critical Heat Flux ◽  
Cold Spray ◽  
Heat Removal ◽  
Spray Coatings ◽  
Spray Process ◽  
External Cooling ◽  
Loss Of Coolant Accident ◽  
Lower Head ◽  
Viable Approach

During a reaction-initiated accident (RIA) or loss of coolant accident (LOCA), passive external-cooling of the reactor lower head is a viable approach for the in-vessel retention (IVR) of Corium; while this concept can certainly be applied to new constructions, it may also be viable for operational systems with existing cavities below the reactor. However, a boiling crisis will inevitably develop on the reactor lower head owing to the occurrence of critical heat flux (CHF) that could reduce the decay heat removal capability as the vapor phase impedes continuous boiling. Fortunately, this effect can be minimized for both new and existing reactors through the use of a cold-spray-delivered, microporous coating that facilitates the formation of vapor microjets from the reactor surface. The microporous coatings were created by first spraying a binary mixture with the sacrificial material then removed via etching. Subsequent quenching experiments on uncoated and coated hemispherical surfaces showed that local CHF values for the coated vessel were consistently higher relative to the bare surface. Moreover, it was observed for both coated and uncoated surfaces that the local rate of boiling and local CHF limit varied appreciably along the outer surface. Nevertheless, the results of this intriguing study clearly show that the use of cold spray coatings could enhance the local CHF limit for downward facing boiling by more than 88%. Moreover, the cold-spray process is amenable to coating the lower heads of operating reactors.

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