scholarly journals Moving From V&V to V&V&C in Nuclear Thermal-Hydraulics

2018 ◽  
Author(s):  
Francesco D’Auria ◽  
Marco Lanfredini
Keyword(s):  
Experimental Data ◽  
Large Diameter ◽  
Nucleate Boiling ◽  
Film Boiling ◽  
Two Phase ◽  
Blow Down ◽  
Best Available Techniques ◽  
Large Heat ◽  
Application Fields ◽  
Bernoulli Flow

V&V constitutes a powerful framework to demonstrate the capability of computational tools in several technological areas. Passing V&V requirements is a needed step before applications. Let’s focus hereafter to the area of (transient) Nuclear Thermal-hydraulic (NTH) and let’s identify V1 and V2 as acronyms for Verification and Validation, respectively. Now, V1 is performed within NTH according to the best available techniques and may not suffer of important deficiencies if compared with other technological areas. This is not the case of V2. Three inherent limitations shall be mentioned in the case of Validation in NTH: 1. Validation implies comparison with experimental data: available experimental data cover a (very) small fraction of the parameter range space expected in applications of the codes; this can be easily seen if one considers data in large diameter pipe, high velocity and high pressure or high power and power density. Noticeably, the scaling issue must be addressed in the framework of V2 which may result in controversial findings. 2. Water is at the center of the attention: the physical properties of water are known to a reasonable extent as well as large variations in values of quantities like density or various derivatives are expected within the range of variation of pressure inside application fields. Although not needed for current validation purposes (e.g. validation ranges may not include a situation of critical pressure and large heat flux) physically inconsistent values predicted by empirical correlations outside validation ranges, shall not be tolerated. 3. Occurrence of complex situations like transition from two-phase critical flow to ‘Bernoulli-flow’ (e.g. towards the end of blow-down) and from film boiling to nucleate boiling, possibly crossing the minimum film boiling temperature (e.g. during reflood). Therefore, whatever can be mentioned as classical V2 is not or cannot be performed in NTH. So, the idea of the present paper is to add a component to the V&V. This component, or step in the process, is called ‘Consistency with Reality’, or with the expected phenomenological evidence. The new component may need to be characterized in some cases and is indicated by the letter ‘C’. Then, the V&V becomes V&V&C. The purpose of the paper is to clarify the motivations at the bases of the V&V&C.

Introducing V&V&C in Nuclear Thermal-Hydraulics

2018 ◽  
Author(s):  
Francesco D’Auria ◽  
Marco Lanfredini
Keyword(s):  
Experimental Data ◽  
Large Diameter ◽  
Nucleate Boiling ◽  
Film Boiling ◽  
Two Phase ◽  
Blow Down ◽  
Best Available Techniques ◽  
Large Heat ◽  
Application Fields ◽  
Bernoulli Flow

V&V constitutes a powerful framework to demonstrate the capability of computational tools in several technological areas. Passing V&V requirements is a needed step before applications. Let’s focus hereafter to the area of (transient) Nuclear Thermal-hydraulic (NTH) and let’s identify V1 and V2 as acronyms for Verification and Validation, respectively. Now, V1 is performed within NTH according to the best available techniques and may not suffer of important deficiencies if compared with other technological areas. This is not the case of V2. Three inherent limitations shall be mentioned in the case of Validation in NTH: 1. Validation implies comparison with experimental data: available experimental data cover a (very) small fraction of the parameter range space expected in applications of the codes; this can be easily seen if one considers data in large diameter pipe, high velocity and high pressure or high power and power density. Noticeably, the scaling issue must be addressed in the framework of V2 which may result in controversial findings. 2. Water is at the center of the attention: the physical properties of water are known to a reasonable extent as well as large variations in values of quantities like density or various derivatives are expected within the range of variation of pressure inside application fields. Although not needed for current validation purposes (e.g. validation ranges may not include a situation of critical pressure and large heat flux) physically inconsistent values predicted by empirical correlations outside validation ranges, shall not be tolerated. 3. Occurrence of complex situations like transition from two-phase critical flow to ‘Bernoulli-flow’ (e.g. towards the end of blow-down) and from film boiling to nucleate boiling, possibly crossing the minimum film boiling temperature (e.g. during reflood). Therefore, whatever can be mentioned as classical V2 is not or cannot be performed in NTH. So, the idea of the present paper is to add a component to the V&V. This component, or step in the process, is called ‘Consistency with Reality’, or with the expected phenomenological evidence. The new component may need to be characterized in some cases and is indicated by the letter ‘C’. Then, the V&V becomes V&V&C. The purpose of the paper is to clarify the motivations at the bases of the V&V&C.

scholarly journals ELETRIC FIELD ENHANCEMENT AND DETERIORATION OF BOILING HEAT TRANSFER AND CRITICAL HEAT FLUX IN DIELECTRIC FLUIDS

2001 ◽  
Vol 1 (1) ◽  
pp. 32
Author(s):  
P. M. Carrica ◽  
V. Masson
Keyword(s):  
Heat Transfer ◽  
Heat Flux ◽  
Critical Heat Flux ◽  
Electric Fields ◽  
Boiling Heat Transfer ◽  
Nucleate Boiling ◽  
Film Boiling ◽  
Two Phase ◽  
Boiling Regime ◽  
Dielectric Fluids

We present the results of an experimental study of the effects of externally imposed electric fields on boiling heat transfer and critical heat flux (CHF) in dielectric fluids. The study comprises the analysis of geometries that, under the effects of electric fields, cause the bubbles either to be pushed toward the heater or away from it. A local phase detection probe was used to measure the void fraction and the interfacial impact rate near the heater. It was found that the critical heat flux can be either augmented or reduced with the application of an electric field, depending on the direction of . In addition, the heat transfer can be slightly enhanced or degraded depending on the heat flux. The study of the two-phase flow in nucleate boiling, only for the case of favorable dielectrophoretic forces, reveals that the application of an electric field reduces the bubble detection time and increases the detachment frequency. It also shows that the two-phase flow characteristics of the second film boiling regime resemble more a nucleate boiling regime than a film boiling regime.

Experimental Study of Well Deviation Correction Model in Deviated and Horizontal Wells Production Profile

2012 ◽  
Vol 490-495 ◽  
pp. 2205-2209
Author(s):  
Jun Feng Liu ◽  
Hai Min Guo
Keyword(s):  
Experimental Data ◽  
Fluid Flow ◽  
Large Diameter ◽  
Horizontal Wells ◽  
Velocity Model ◽  
Data Interpretation ◽  
Internal Diameter ◽  
Affecting Factors ◽  
Vertical Wells ◽  
Two Phase

There are big difference of fluid flow patterns between horizontal wells and vertical wells, so the current interpretation models of production logging multiphase flow in vertical wells are not suitable for data interpretation in highly deviated and horizontal wells. In this paper, firstly, the two-phase flow (oil-water and gas-water) simulation experiments have been carried out in large-diameter (0.124 meter internal diameter) uphill, horizontal and downhill Plexiglas pipe with practical production logging tools. Secondly, based on the conclusions of fluid flow mechanism from experimental data analysis, and considering the affecting factors (i.e. Logging tool and well deviation ), we have obtained slip velocity model after well deviation correction in highly deviated and horizontal wells, which have been corrected by the mature interpretation models. Finally, this proposed method has been proved correct and feasible through the experimental data validation.

Strategy for the Development of a DNB Local Predictive Approach Based on Neptune CFD Software

2006 ◽  
Author(s):  
Pierre-Antoine Haynes ◽  
Pierre Pe´turaud ◽  
Michae¨l Montout ◽  
Eric Hervieu
Keyword(s):  
Experimental Data ◽  
Two Phase Flow ◽  
Bubbly Flow ◽  
Nucleate Boiling ◽  
Phase Flow ◽  
Thermal Hydraulics ◽  
Two Phase ◽  
Validation Data ◽  
Simulation Tools ◽  
Predictive Approach

The NEPTUNE project constitutes the thermal-hydraulics part of a long-term joint development program for the next generation of nuclear reactor simulation tools. This project is being carried through by EDF (Electricite´ de France) and CEA (Commissariat a` l’Energie Atomique), with the co-sponsorship of IRSN (Institut de Radioprotection et de Suˆrete´ Nucle´aire) and AREVA NP. NEPTUNE is a multi-phase flow software platform that includes advanced physical models and numerical methods for each simulation scale (CFD, component, system). NEPTUNE also provides new multi-scale and multi-disciplinary coupling functionalities. This new generation of two-phase flow simulation tools aims at meeting major industrial needs. DNB (Departure from Nucleate Boiling) prediction in PWRs is one of the high priority needs, and this paper focuses on its anticipated improvement by means of a so-called “Local Predictive Approach” using the NEPTUNE CFD code. We firstly present the ambitious “Local Predictive Approach” anticipated for a better prediction of DNB, i.e. an approach that intends to result in CHF correlations based on relevant local parameters as provided by the CFD modeling. The associated requirements for the two-phase flow modeling are underlined as well as those for the good level of performance of the NEPTUNE CFD code; hence, the code validation strategy based on different experimental data base types (including separated effect and integral-type tests data) is depicted. Secondly, we present comparisons between low pressure adiabatic bubbly flow experimental data obtained on the DEDALE experiment and the associated numerical simulation results. This study anew shows the high potential of NEPTUNE CFD code, even if, with respect to the aforementioned DNB-related aim, there is still a need for some modeling improvements involving new validation data obtained in thermal-hydraulics conditions representative of PWR ones. Finally, we deal with one of these new experimental data needs and present a scaling method for the design of the associated experimentation devoted to the analysis of the dynamics-related modeling of a bubbly flow in PWR representative conditions.

The Relation Between Bubble Diameter and Frequency of Removal From a Sphere During Film Boiling

1966 ◽  
Vol 88 (1) ◽  
pp. 87-90 ◽  
Author(s):  
T. H. K. Frederking ◽  
D. J. Daniels
Keyword(s):  
Experimental Data ◽  
Liquid Nitrogen ◽  
Bubble Diameter ◽  
Nucleate Boiling ◽  
Film Boiling ◽  
Saturated Liquid ◽  
Bubble Frequency

The kinematics of vapor removal from a sphere has been investigated during film boiling in saturated liquid nitrogen. New experimental data are presented concerning the bubble frequency f and diameter D. In spite of film boiling conditions, the results may be correlated with the nucleate boiling relation, f(D)1/2 ≈ 17.5 (cm)1/2 sec, proposed by McFadden and Grassmann.

scholarly journals ELETRIC FIELD ENHANCEMENT AND DETERIORATION OF BOILING HEAT TRANSFER AND CRITICAL HEAT FLUX IN DIELECTRIC FLUIDS

2002 ◽  
Vol 1 (1) ◽  
Author(s):  
P. M. Carrica ◽  
V. Masson
Keyword(s):  
Heat Transfer ◽  
Heat Flux ◽  
Critical Heat Flux ◽  
Electric Fields ◽  
Boiling Heat Transfer ◽  
Nucleate Boiling ◽  
Film Boiling ◽  
Two Phase ◽  
Boiling Regime ◽  
Dielectric Fluids

We present the results of an experimental study of the effects of externally imposed electric fields on boiling heat transfer and critical heat flux (CHF) in dielectric fluids. The study comprises the analysis of geometries that, under the effects of electric fields, cause the bubbles either to be pushed toward the heater or away from it. A local phase detection probe was used to measure the void fraction and the interfacial impact rate near the heater. It was found that the critical heat flux can be either augmented or reduced with the application of an electric field, depending on the direction of . In addition, the heat transfer can be slightly enhanced or degraded depending on the heat flux. The study of the two-phase flow in nucleate boiling, only for the case of favorable dielectrophoretic forces, reveals that the application of an electric field reduces the bubble detection time and increases the detachment frequency. It also shows that the two-phase flow characteristics of the second film boiling regime resemble more a nucleate boiling regime than a film boiling regime.

Heat Transfer in Vapour-Liquid Flow at High Reduced Pressures

2010 ◽  
Author(s):  
Victor Yagov ◽  
Maria Minko
Keyword(s):  
Heat Transfer ◽  
Experimental Data ◽  
Boiling Heat Transfer ◽  
Flow Boiling ◽  
Nucleate Boiling ◽  
Phase Flow ◽  
Two Phase ◽  
Co2 Flow ◽  
Reduced Pressures ◽  
Good Agreement

During the last decade a number of studies of boiling heat transfer in carbon dioxide notably increase. As a field of CO2 practical using corresponds to high reduced pressures, and a majority of available experimental data on CO2 flow boiling even in submillimetric channels relate to turbulent liquid flow regimes, a possibility arises to develop sufficiently general method for HTC predicting. Under the above conditions nucleate boiling occurs up to rather high flow quality, even in annular flow regime due to extremely small size of an equilibrium vapour bubble. This conclusion is in agreement with the available experimental data. The predicting equation for nucleate boiling heat transfer developed by one of the present authors in 1988 is valid for any nonmetallic liquid. A contribution of forced convection in heat transfer is calculated according to the Petukhov et al. equation with correction factor, which accounted for an effect of velocity increase due to evaporation. This effect can be essential at relatively small heat fluxes and rather high mass flow rates. The Reynolds analogy and homogeneous model are used in order to account for the convective heat transfer augmentation in two-phase flow. Due to low ratio of liquid and vapour densities at high reduced pressures the homogeneous approximation of two-phase flow seems to be warranted. A total heat transfer coefficient is calculated as an interpolated value of boiling and convective HTCs. The experimental data on CO2 flow boiling related to regimes before heated wall dryout incipience are in rather good agreement with the calculations. Besides the data on carbon dioxide flow boiling, the results on water, helium, nitrogen and some refrigerants were used for comparison; at rather high reduced pressures the computed and the measured values of HTCs are in a good agreement. The data include results obtained in the channels of a diameter from 0.6mm up to 18mm. It is clear that at high reduced pressures there is no strong variation in boiling heat transfer with channel size decrease, it means that a classification on channel size has no sense if it does not consider liquid/vapour densities ratio.

Experimental Investigation of Two-Phase Heat Transfer in a Simulated Cooling Duct of a Piston Engine Cylinder Head

2020 ◽  
pp. 4-15
Author(s):  
O.V. Abyzov ◽  
Yu.V. Galyshev ◽  
A.K. Ivanov
Keyword(s):  
Heat Transfer ◽  
Experimental Data ◽  
Experimental Investigation ◽  
Cylinder Head ◽  
Single Phase ◽  
Nucleate Boiling ◽  
Piston Engine ◽  
Two Phase ◽  
Engine Cylinder ◽  
Two Phase Heat Transfer

Liquid cooling of cylinder and piston parts in highly boosted internal combustion engines is generally accompanied by local phase transition phenomena, such as surface nucleate boiling. The heat transfer coefficient of nucleate boiling is several times higher than that of single-phase convection. In order to efficiently exploit the thermal effect of nucleate boiling in cooling systems, simultaneously preventing emergency supercritical modes, a deeper understanding of boiling physics based on full-scale experiments is required. We conducted experimental investigation of heat transfer in a simulated cooling duct of a piston engine cylinder head, using a bespoke motor-free installation. We studied the effects of velocity, flow character and coolant type on the heat transfer, accounting for the presence of congestion regions. Over the course of the experiment, we simulated thermal conditions characteristic of different heat transfer types: single-phase convection, nucleate boiling, the onset of boiling crisis. We used the experimental data to plot the coolant heat flow density as a function of wall temperature for different measuring points situated inside the stream and the turbulent flow regions (congestion regions). We show that the mature nucleate boiling mode is the most favourable in terms of how uniform the temperature field within a part is. The experimental data obtained during the investigation may be used to verify mathematical simulations in the two-phase heat transfer theory, provided the data have been appropriately processed

Forced Convection Film Boiling Heat Transfer From Single Horizontal Cylinders in Saturated and Subcooled Liquids: Part 1 — Experimental Data and Its Correlation for Saturated Liquids

2008 ◽  
Author(s):  
Qiusheng Liu ◽  
Katsuya Fukuda
Keyword(s):  
Heat Transfer ◽  
Experimental Data ◽  
Forced Convection ◽  
Boiling Heat Transfer ◽  
Horizontal Cylinder ◽  
Film Boiling ◽  
Two Phase ◽  
Saturated Water ◽  
Horizontal Cylinders ◽  
Flow Velocities

Forced convection film boiling heat transfer on a horizontal cylinder in saturated water and Freon-113 flowing upward perpendicular to the cylinder was measured for the flow velocities ranging from zero to 1 m/s at the system pressures ranging from 100 to 500 kPa: the platinum cylinders with the diameters ranging from 0.7 to 5 mm were used as the test cylinder heaters. The existing correlation for forced convection film boiling heat transfer given by Bromley et al. could not well describe the experimental data obtained, especially those for the higher pressures. The forced convection film boiling heat transfer correlation including the radiation contribution from the cylinders with various diameters for saturation conditions was developed based on forced convection two-phase laminar boundary layer film boiling model and the experimental data obtained. The experimental data agreed with the corresponding values derived from the correlation within ±15% for the flow velocities below 0.7 m/s, and within −30% to +15% for higher flow velocities. It was confirmed that the experimental data obtained by Bromley et al. for the horizontal carbon cylinders with the diameters ranging from 9.83 to 16.2 mm and with the significant radiation effect from the cylinder surfaces in various liquids for the various flow velocities up to 4.4 m/s at an atmospheric pressure agreed with the corresponding values derived from the new correlation within ±20%.

scholarly journals HEAT TRANSFER MODELLING OF TWO-PHASE FLOW IN ISOLATED CHANNEL

2018 ◽  
Vol 14 ◽  
pp. 49
Author(s):  
Daniel Vlček ◽  
Vojtěch Caha ◽  
Martin Ševeček
Keyword(s):  
Heat Transfer ◽  
Nucleate Boiling ◽  
Film Boiling ◽  
Phase Flow ◽  
Two Phase ◽  
Boiling Crisis ◽  
Simple Geometry ◽  
Inverted Annular Film Boiling ◽  
Transfer Modelling ◽  
Annular Film

This paper deals with Post-CHF (critical heat flux) heat transfer with the focus on different regimes of film boiling. The new thermal-hydraulic code TUBE 2.0 is presented. This code uses the equation of energy conservation and predefined correlations to establish wall temperature, the departure of nucleate boiling ratio as well as other parameters of cooling in a simple geometry - an isolated channel. With experimental data of inverted annular film boiling from Stewart, the best-performing correlation for calculation of post-CHF heat transfer in the channel was determined. Finally, the new presented code TUBE 2.0 and subchannel code SUBCAL owned by Chemcomex a.s. are compared using results of various experiments conducted by Becker. Data from Stewart could not be used because of inability to predict the onset of boiling crisis with several correlations.

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