The NURESIM Project of the 6th European Framework Program initiated the development of a new-generation common European Standard Software Platform for nuclear reactor simulation. The thermal-hydraulic subproject aims at improving the understanding and the predictive capabilities of the simulation tools for key two-phase flow thermal-hydraulic processes such as the critical heat flux (CHF). As part of a multi-scale analysis of reactor thermal-hydraulics, a two-phase CFD tool is developed to allow zooming on local processes. Current industrial methods for CHF mainly use the sub-channel analysis and empirical CHF correlations based on large scale experiments having the real geometry of a reactor assembly. Two-phase CFD is used here for understanding some boiling flow processes, for helping new fuel assembly design, and for developing better CHF predictions in both PWR and BWR. This paper presents a review of experimental data which can be used for validation of the two-phase CFD application to CHF investigations. The phenomenology of DNB and Dry-Out are detailed identifying all basic flow processes which require a specific modeling in CFD tool. The resulting modeling program of work is given and the current state-of-the-art of the modeling within the NURESIM project is presented.
The NURESIM Integrated Project of the 6th European Framework Programme is envisaged to provide the initial step towards a common European Standard Software Platform for modeling, recording, exchanging, and recovering data for nuclear reactors simulations. Key objectives of NURESIM include the integration of advanced physical models in a shared, open software platform, incorporating the latest advances in reactor core physics, thermal hydraulics, and coupled multiphysics modeling. The specific objectives of NURESIM are to initiate the development of the next generation of experimentally validated, “best-estimate” tools with improved prediction capabilities, standardization, and robustness to address current and future needs of industry, reactor safety organizations, academic, government, and private institutions.
The overall
objective of NURESIM thermal-hydraulic subproject is to improve the understanding
and the predictive capabilities of the simulation tools for key two-phase flow
thermal-hydraulic processes that can occur in nuclear reactors, focusing on two
high priority issues, the critical heat flux (CHF), and the pressurized thermal
shock (PTS). This overall objective has resulted from the conclusions of the
EUROFASTNET [
This paper presents a review of existing experimental data bases which can be used for validation of the two-phase CFD application to critical heat flux (CHF) investigations with respect to nuclear reactors. The phenomenology of DNB and dryout is detailed identifying all basic flow processes which require a specific modeling in CFD tool. The resulting programme of work is given, and the current state of the art of the modeling is presented.
Four basic spatial scales encountered in thermal-hydraulic phenomena relevant to nuclear power plants:
system scales, which are addressed
by zero- and one-dimensional flow models for pipes, pumps, valves, breaks, and
control systems together with CFD methods for porous media; component-scales, which are
addressed by CFD methods for porous media (typically for the core of a reactor
or for the steam generators with a minimum spatial resolution in the case of
the subchannel analysis);
mesoscales, which are addressed by
computational fluid dynamics (CFDs) methods in open medium, including
turbulence models, using either Reynolds-averaged simulations (RANSs) or large
eddy simulation (LES);
microscales, which are addressed
by direct numerical simulation (DNS) and interface tracking methods (ITMs) that
focus on a very small domain (e.g., a domain containing a few bubbles or
droplets).
In CHF investigations, the present industrial methods mainly use the component scale with 3D modeling of core assemblies using in the hot assembly the subchannel analysis. Large-scale experiments having the real geometry of the reactor assembly are used to develop empirical correlation for the CHF as function of flow variables which are averaged over the cross-section of a subchannel. The NURESIM-TH activities regarding CHF aim at using two-phase CFD as a tool for understanding boiling flow processes, in order to subsequently help new fuel assembly design and to develop better CHF predictions in both PWR and BWR. A “local predictive approach” may be envisaged for the long term where CHF correlations would be based on local (mesoscale) T/H parameters provided by CFD. If the processes leading to DNB and dryout are well understood, the CHF correlation will be physically based, but one may also develop empirical correlations if some phenomena are not clearly identified.
Considering the rather low maturity
of two-phase CFD, a general methodology was proposed by a Writing Group of the
OECD-CSNI (see Bestion et al. 2006 [
identification of all important
flow processes of the application,
selecting a basic model,
filtering turbulent scales and
two-phase intermittency scales,
identification of local interface structure,
modeling interfacial transfers,
modeling turbulent transfers,
modeling wall transfers,
use
of finer scale simulations for modeling,
identification of validation and verification
test cases with possibly some demonstration test cases.
The choice of a validation test matrix and of the basic modeling approach should be consistent with each other since there must be enough measured physical parameters to be able to validate separately each sensitive process modeled in the equations.
The identification of the basic flow processes related to both DNB and dryout and a review of available experimental data were performed before selecting a basic model and defining a development and validation programme. Next sections will present this initial work and will conclude on the present state of the art in the modeling within the NURESIM project.
Departure from nucleate boiling is the main governing critical heat flux mechanism for pressurized water reactors. A huge amount of work has been devoted to the DNB in the past decades but the evaluation of the CHF still relies on fully empirical methods.
Rod bundles with spacer grids are tested in real conditions with the fuel assembly geometry and the same flow T/H conditions as in the reactor. Such experiments are very expensive and time consuming but necessary to determine the CHF behaviour of any new fuel assembly design.
The reason of this situation is that the phenomenology of convective boiling and DNB is very complex, and many small-scale processes are not well understood. It is very likely that phenomena occurring at various scales play a role; one can distinguish three scales for reactor DNB phenomenology.
The
The
A nonexhaustive list of flow processes at the various scales is given here below.
Activation of nucleation sites. Evolution of active sites density
with increasing power. Growing of attached bubbles. Sliding of attached bubbles along
heating wall. Coalescence of attached bubbles. Extension of dry patch.
Effects of wall conductivity and heat capacity. Detachment of bubbles. Rewetting after detachment. Mutual influence of neighboring
nucleation sites. Influence of flow characteristics on
local processes: external convective velocity. Behaviour of detached bubbles:
coalescence, migration. Interactions between detached
bubbles. Forces between detached bubbles and
liquid flow. Formation of high-void layer if
bubbles cannot escape due to counter current flow limitation (CCFL) type
phenomenon and behaviour of the thin liquid film which vaporizes below the
bubble layer.
Wall to fluid heat transfer in
subcooled boiling: liquid heating, vaporization, quenching. Transport and dispersion of bubbles. Vaporization-condensation of
bubbles. Coalescence and breakup of bubbles. Turbulent transfers of heat and momentum
within liquid. Effects of polydispersion of bubbles
on interfacial transfers Local effects of grids: enhanced
turbulence and flow rotation.
Mixing between subchannels,
cross-flows, turbulence. Grid spacers effects on mixing
between sub-channels. Effects of cross-sectional averaged pressure
Effects of nonuniform heat flux on
DNB occurrence. Effects of spacer grids on DNB
occurrence.
Two-phase CFD predictions should be compared to relevant experimental data in order to validate all mesoscale flow processes, in geometrical and T/H conditions preferably representative of the industrial ones. This will bring a better understanding of the effects of the mesoscale phenomena on the CHF occurrence. Moreover, microscale flow phenomena should also be better understood for developing physically based closure laws in the CFD approach. In this purpose, any experimental information on such microscale phenomena or any DNS simulations may be used to improve the CFD simulation tool. However, this project did not bring enough information to build a physically based DNB criterion. Nevertheless, CFD simulations of boiling flowup to DNB have the potentiality to predict some mesoscale effects on flow conditions at the wall such as the development of two-phase boundary layers, or spacer grid effects, which are not seen by the subchannel analysis and current empirical CHF models. One may at least expect that the effects of nonuniform axial heat flux, which are now empirically modeled, may be simply seen by local conditions resulting from CFD predictions. Also the effects of spacer grid design on flow conditions seen by the wall may be described at the CFD scale whereas subchannel analysis can only describe the associated pressure loss, the additional mixing between neighboring subchannels and the effect on CHF when experimental data are available.
The following data sources were
reviewed and analysed with respect to their interest for validating CFD tools
used in DNB investigations. Table
There are single phase liquid data (AGATE) which may be used as a first step in the validation of turbulence models in a rod bundle with spacer grids. Some air-water bubbly flow data (DEDALE, TOPFLOW) may be used as a first step in the validation of models for bubble transport and dispersion, coalescence and breakup, effects of polydispersion on interfacial forces, and momentum turbulent transfers. Boiling flow data in simple geometry (DEBORA, ASU, Purdue data, KAERI data) either in steam-water of Freon (R12, R113) may then be used to further validate in more representative conditions (pressure is either atmospheric or similar to reactor conditions) the models already investigated in air-water conditions, with additional effects of wall heat transfers, turbulent heat transfers and interfacial heat, and mass transfers due to vaporization and condensation. Some DEBORA data were recorded in conditions which were very close to CHF occurrence. Effects of spacers are also validated in boiling flow conditions with the DEBORA-Promoter data. BFBT data are used to validate the void distribution of a steam-water boiling flow in a real BWR rod bundle geometry. These data are unique and can also be used to some extent for DNB investigations if one considers the low quality data. LWL data in a real-rod bundle of a WWER reactor finally allow a global validation of the boiling flowup to DNB.
DEDALE is an adiabatic air-water two-phase
experimental programme performed at EDF/DER [
The DEBORA experiment [
The test section is an electrically heated vertical tube with upward R12 boiling flow simulating PWR in-core T/H conditions, with local measurements along a diameter within the outlet tube cross section of both steam phase characteristics (void fraction, interfacial area concentration, bubble size, and mean axial velocity) and liquid phase parameter (temperature).
The “DEBORA-Promoter” tests (see Figure
“DEBORA-promoter” geometry.
Validation of CFD tools on these
tests provides additional information on the effect of spacer grid wake on the
mixing of bubbles generated at the wall and on the effects of the flow rotation
on the void repartition; simulations of such tests with Neptune_CFD were
presented [
The AGATE experiment has been developed in CEA Grenoble. Two-test sections were used:
“AGATE-Grid” consists of a
“AGATE-Promoter” with a similar
geometry as “DEBORA-Promoter” one (i.e., pipe with a 3-vane turbulence
enhancer).
Nonheated water flows upwardly in the vertical test section, and velocity measurements are made using laser Doppler anemometry (LDA). Both the mean velocity and velocity fluctuations are measured in order to investigate the effects of the grid or promoter.
The data allow to validate the
turbulence modeling with spacer grid (or turbulence promoter/enhancer) effects
in single-phase conditions. They were used for validation of a 1D model with
QLOVICE tests are being performed by CEA in order to investigate basic processes associated with DNB. QLOVICE is a visualization of pool boiling with high-speed video-camera.
A transparent heated bottom wall
allows to visualise the bubble nucleation and detachment.
A side window allows to see bubble
behaviour after detachment.
First tests were performed and have clearly shown the dry patch evolutions. It was observed
bubble sliding along the heating
wall before detachment,
sudden large size dry patch
extension observed followed by a wall rewetting,
many bubble clusters,
interactions between neighbouring
nucleation sites.
Two main processes are assumed to play a significant (dominant) role on the DNB occurrence: a sudden extension of dry patch up to DNB or a CCFL type phenomenon with bubbles which cannot escape from wall after detachment. However, no conclusion can be presently drawn on the dominant process.
Experiments of turbulent subcooled
flow in a vertical annular channel were carried out at the Arizona State University
[
Validation of CFD tools on ASU tests provides information on the steam production at the wall in subcooled boiling, on the interfacial forces responsible for the void profiles, on interfacial heat and mass transfers, on interfacial area concentration evolution, and on turbulence in the bubbly boundary layer.
Measurements used simultaneously a two-component laser Doppler velocimetry for liquid velocity and a fast response cold-wire for temperature field, as well as a dual-sensor fiber optic probe for the vapour fraction and vapour axial velocity.
A comparison of Neptune simulations with the early tests was presented in [
Experiments have been carried out
at the School of Nuclear Engineering of Purdue University in an internally
heated annulus to provide local measurements of void fraction, interfacial area
concentration, and interfacial velocity in subcooled boiling [
Earlier tests [
Visual observations of the boiling
processes provided essential information on the displacement between the
location of net vapor generation (NVG) and the location of bubble detachment [
A few analyses to test the validity
of CFD codes have been carried out using the earlier series of test data [
Experiments have been carried out
at the Korea Atomic Energy Research Institute (KAERI) in an internally heated
annulus to provide local measurements of void fraction and phase velocities in
subcooled boiling [
Measurements of void fraction and
bubble velocity were taken using a double-sensor conductivity probe. Liquid
velocities were measured by a Pitot tube, correcting for the effect of bubbles
[
Tests have been used for assessing
the CFX-4 code [
The structure of an adiabatic air-water and of
steam-water flow with reduced condensation and with slight subcooling in a
vertical pipe of 195.3 mm inner diameter (DN200) was studied using wire-mesh
sensors. The experiments were performed at the two-phase FLOW test
facility (TOPFLOW) [
Function and construction of wire-mesh sensors are
described in [
A technique to analyse the evolution of the flow
structure is the calculation of radial gas fraction profiles decomposed
according to bubble size classes [
A visualisation (see Figure
Virtual side projections (left halves of the columns) and side
views of virtual central cuts (right halves) of the mesh-sensor data (from
[
The data can be used to test the complex interaction
of local bubble distributions, bubble size distributions, and local heat
and mass transfer. The lateral motion of the bubbles in a shear flow, bubble
coalescence, and
breakup and the phase transfer can be observed by measurements along the pipe. For
example, the radial distribution of bubbles strongly depends on their diameter.
For a vertical cocurrent upwards flow,
smaller bubbles tend to move towards the wall, while large bubbles are
preferably found in the centre. Details on the steam-water experiments and
investigations on the modeling of such flows are presented by Lucas and Prasser [
Experimental tests for measuring the void fraction distribution inside boiling
water reactor (BWR) fuel assemblies have been conducted by the Nuclear Power
Engineering Corporation (NUPEC), Tokyo,
Japan, by the use of an experimental facility
referred to as BFBT (BWR Full-size Fine-mesh Bundle Tests). Data provided by
such facility have been initially used for subchannel code assessment [
The test loop has a full range of steady-state void fraction testing capabilities over BWR operating conditions. Unsteady characteristics, flow changes, power changes, and complicated BWR operational transients are simulated too.
The test section consists of a full-scale BWR fuel assembly simulator, which is made of electrically heated rods able to reproduce the actual power profiles generated by nuclear fission. The instrumentation allows measurements of temperature, flow rate, pressure and, mainly, void fraction.
An X-ray CT scanner, consisting of an X-ray tube and 512 detectors, is employed to measure
the void fraction in the upper part of the test section in steady-state
conditions. The void fraction data have a
The large water loop has been built
at the NUCLEAR MACHINERY PLANT,
The test sections were formed by 7 or 19 parallel electrically heated rods with external diameters of 9 mm. Axial and radial uniform or nonuniform heat flux distribution and water up flow were used in the tests. The rods were with direct heating were specially manufactured with axially varying wall thickness while maintaining a constant outside diameter to achieve nonuniform axial heat flux. The rods (3500 mm long) were placed in regular hexagonal geometry with a pitch of 12.5–13 mm. Critical conditions were obtained under constant thermal-hydraulic conditions by gradually increasing heat input.
Based on data and manpower
availability, the following programme of validation was planed to be performed
within the NURESIM project (see Table
Planed validation and demonstration calculations within NURESIM project.
Validation tests and demonstration tests | Validation Demonstration | Main interest of validation |
---|---|---|
DEBORA | V | Investigations of wall heat transfer models |
DEBORA tests close to CHF conditions | V | Looking for processes responsible for void accumulation close to the wall |
DEBORA and/or TOPFLOW polydispersed data | V | Validation of the method of statistical moments |
TOPFLOW polydispersed data | V | Validation of the MUSIG method |
DEBORA polydispersed data | V | Validation of the MUSIG method |
DEDALE | V | Evaluation of LES simulation of bubbly flow |
ASU boiling water experiment | V | Validation of wall function Evaluation of LES simulation of boiling bubbly flow |
BFBT experiments | VD | Evaluation of models controlling void distribution in actual core geometry |
DEBORA | V | Investigations of wall heat transfer models |
Large water loop (LWL) | VD | Evaluation of CHF prediction CFD in actual core geometry |
DEDALE | DEBORA | DEBORA Promoter | AGATE (&AGATE-promoter) | ASU | PURDUE boiling | KAERI boiling | TOPFLOW | BFBT | LWL | |
---|---|---|---|---|---|---|---|---|---|---|
Geometry | Tube | Tube
| Tube + promoter | Rod bundle (tube +promoter) | Annulus
| Annulus | Annulus
| Tube
| Rod bundle | Rod bundle |
Fluid/flow conditions | Air-water adiabatic | R12 | R12 | water | R113 | Steam-water | Steam-water | Air-water and steam-water | Steam/ water | Steam/ water |
Local measur-ements | ||||||||||
Wall to fluid heat | X | X | X | X | ||||||
Bubbles transport and dispersion | X | X | X | X | X | X | ||||
Vaporization-condensation | X | X | X | X | X | |||||
Coalescence and breakup of bubbles | X | X | X | X | X | X | ||||
Turbulent Transfers of heat and momentum | X | X | X | X | X | X | X | X | ||
Effects of polydispersion | X | X | ||||||||
Effects of grids | X | X | ||||||||
Combined effects in real geometry | BWR | PWR WWER |
Table
The present data basis is not sufficient to validate all phenomena of interest, and the main defaults are the lack of turbulence data in high void bubbly flow and the lack of data for validation of the heat flux partitioning at the wall in convective nucleate boiling. More generally no data can provide information on microscale phenomena which makes the development of physically based models in the near wall region difficult.
The following state of the art on the modeling of two-phase flow up to DNB occurrence results from the ongoing work in NURESIM which mainly addressed flow conditions before DNB.
(1) Basic model: as boiling bubbly flows are encountered, the two-fluid model is naturally used in this flow conditions to benefit from the possibility to model all interfacial forces acting on the bubbles such as drag, lift, turbulent dispersion, virtual mass, and wall forces which control the void repartition in a boiling channel. The choice of the method to model poly-dispersion effects remains partly open.
(2)
Averaging or filtering equations: considering flow in a PWR core in
conditions close to nominal, when boiling occurs, a high velocity steady flow
regime takes place with times scales associated to the passage of bubbles being
very small (
(3) Identification of local interface structure: there is a unique interfacial structure corresponding to a dispersed gas phase in a continuous liquid. As long as bubbly flow is encountered, there is no need to develop an identification of the local flow regime and there is no need to use an ITM. Going to DNB occurrence, a gas layer appears and a criterion must be implemented for identifying this occurrence. A very simple criterion based on the local void fraction was applied to LWL tests. However, the description of the interface structure may require addition of transport equations such as interfacial area transport (IAT) or bubble number density transport. More generally, the method of the statistical moments (MMSs) can be used to characterise the poly-dispersion of the vapour phase with a bubble size spectrum. Another approach of the poly-dispersion is to use a multigroup model (MUSIG method) with mass (and momentum) equations written for several bubble sizes. These two methods are being used, evaluated, and compared on both DEBORA and TOPFLOW tests. The MUSIG method with several mass equations for different bubble sizes and at least two momentum equations has shown good capabilities for capturing all qualitative effects in TOPFLOW vertical pipe tests. The MMS has been applied to a subcooled boiling DEBORA test, demonstrating a significant effect of polydispersion on the condensation predictions. MUSIG and MMS still have to be further evaluated.
(4) Momentum transfer control the void distribution and it is necessary to model all the forces acting on the bubbles. The virtual mass force is not expected to play a very important role, and rather reliable models exist for the drag force. More effort should be paid to the modeling and validation of both lift and turbulent dispersion forces since available models are still often tuned. In particular, since the lift force may depend on the bubble size, it is now necessary to model poly-dispersion to take this into account.
(5)
Turbulent transfers: liquid turbulence plays a very important role in
boiling flows. It influences liquid temperature diffusion, bubble dispersion, bubble
detachment, bubble coalescence, and breakup which affect the interfacial area. Then,
the liquid turbulent scales have to be predicted correctly to model all these
processes and this will require additional transport equations. The
(6)
Wall-to-fluid transfers: modeling of velocity profiles in the near-wall
boiling region was improved by implementing the two-phase wall function in
momentum equations. Models were validated on ASU boiling flow tests [
(7) First demonstration test cases were performed with Neptune_CFD calculations of critical heat flux tests in the LWL loop which is prototypical of WWER type core assemblies. Computational grid consists of 150 000 hexahedral cells. Although the simulation is not fully successful quantitatively, Neptune showed the capability to model boiling flow in a complex industrial geometry and in reactor flow conditions up to CHF. CHF occurrence was predicted at the right location but with errors from 1% to 25% on the heat flux, which shows how far we still are from the final goal of the “local predictive approach.”
Annular
flow pattern usually is the predominant flow regime in upper core regions in boiling
water reactors. The limitation of the total power obtained from each assembly
is the occurrence of dryout. Increasing the heat flux above some critical value
can lead to dryout that is associated with a sudden increase in the wall
temperature, which, in turn, can destroy the cladding material and allow the
radiation releases into the primary system. The phenomenology of dryout in
annular mist flow was described in [
The liquid phase exists as a liquid film, which is attached to walls, and as droplets, which are carried in the central part of the channel by the vapour phase.
The mass flow rate in the liquid film is changing due to several mass transfer mechanisms.
Due to hydrodynamic forces acting on
the liquid film surface, certain amount of liquid from liquid film is entrained
into the vapour core.
Another mechanism that is causing
liquid film depletion is associated with evaporation due to heating applied to
walls.
These two mechanisms must be
counterbalanced by drop deposition from the vapour core to the liquid film
surface to avoid film dryout.
There
are several possible mechanisms that have been postulated for dryout (Hewitt,
1982 [
The liquid film dries by progressive
entrainment and evaporation, which are prevailing in comparison to deposition,
and dryout occurs when the film has gone.
Formation of a dry patch within the
liquid film, causing such wall temperature increase that cannot be rewetted. In
some situations a sudden disruption of liquid film may occur beyond which the wall
surface is dry. The disruption mechanism is not fully understood yet, however,
hydrodynamic mechanisms for the disruption are postulated.
For very thin liquid films. dryout
occurs when the rate of evaporation at the surface exceeds the rate at which droplets
arrive at the surface due to deposition.
For thicker liquid films, it is
postulated that dryout may occur due to vapour film formation under the liquid
film. The mechanism of forming this vapour film might be of the same type as
described for the DNB mechanisms.
Annular regime in boiling flow is characterized by a thin liquid film flowing on the channel walls and a gas core flowing in the central part of the channel. The droplets in the gas core represent a larger interfacial area than the liquid film and thus can dominate heat and mass transport between the phases. System pressure drop is increased by droplet acceleration in the gas core, and depositing droplets contribute to corrosion by increasing local wall friction.
To some extent, the dryout is a more simple process than the DNB since one cannot list so many microscale phenomena which may play a role. In particular, if one first focuses on the first dryout scenario with entrainment and evaporation prevailing in comparison to deposition, only mesoscale phenomena have to be considered.
The most important mesoscale phenomena and parameters in annular flow affecting the occurrence of dryout are
drop size, deposition of droplets, entrainment of droplets, and film thickness.
Drop
size is an important parameter which affects the deposition rates and thus the
dryout phenomenon. It can be described by a size PDF,
Liquid droplets carried by a turbulent gas stream will deposit on bounding walls. Clearly, deposition rate will have an important influence on the dryout occurrence.
It may depend on several unresolved issues, such as turbulence-particle interactions and drop breakup and coalescence.
Deposition rate will depend on drop dispersion in turbulent flow where particle motion is primarily governed by interactions with eddies of various scales. Depending on the ratio of the particle response time to the eddy characteristic time, the dispersion can have different characters. If this ratio is very small, particles are following the continuous flow structure. When the ratio is close to 1 (the time constants of eddies and particles are of the same range of magnitude), the dispersion of drops can be even bigger than that observed in the carrier fluid. Finally, for high values of the ratio particles remain largely unaffected by eddies.
Typically, drop deposition is associated with two mechanisms: the diffusion process and the free-flight to the wall. For proper prediction of the deposition rate of droplets, both these mechanisms have to be taken into account. In addition, impinging conditions of a drop on a liquid surface have to be considered. When a single droplet impinges a liquid film, various phenomena can occur. The droplet can bounce from the surface or merge with the liquid film. Splash can occur when the drop kinetic energy is high enough. For conditions typical for BWRs, the liquid film is thin and the velocity of droplets is high, thus splashing and mergence are the key phenomena involved.
Several mechanisms of drop entrainment from the liquid film have been identified. The dynamic impact of gas core causes generation of waves on the film surface, with droplets being separated and entrained from the crests of these waves. The creation and breakup of the disturbance waves play important roles in the drop entrainment process. Another entrainment mechanism is associated with splashing associated with drop deposition, as already mentioned in the previous section. Finally, in a heated channel with nucleate boiling in the film, entrainment can occur due to the action of vapour bubbles which induce splashing.
Calculation of the liquid film thickness is necessary to predict the occurrence of dryout. To obtain the liquid film thickness and velocity, it is necessary to solve the mass and momentum conservation equations of the film in order to determine the film flowrate and pressure drop. This requires proper modeling of deposition, entrainment, and evaporation in mass equation and of the wall friction and interfacial friction in the momentum equation which depend on the wave structure of the film interface.
Early
experiments were focused on the measurements of the total power, which was
necessary for the dryout occurrence in a heated channel. A vast number of these
experiments were performed for different conduit geometries in different flow
conditions. The measurements for steam-water were done in round ducts, annuli,
and rod clusters. Measurements in annuli covered the pressures of 30, 50, and
70 bar (Becker and
Letzer [
Würtz [
An extensive review of existing measurements of deposition
rate has been presented by Okawa et al. [
It was experimentally proven that the mode of the
deposition is dependent on the droplet size. Observations of droplet motion
(Andreussi [
One way to measure entrainment is to reach a
quasiequilibrium state in the system where it is considered that deposition
rate is equal to the entrainment rate. Okawa et al. [
Table
Reference | Measured value | Geometry | Fluid/heating | Flow conditions |
---|---|---|---|---|
Würtz, 1978 [ | Tube test section:
| Steam-water adiabatic and diabatic | ||
Andreussi, 1983 [ | Plexiglass tube
| Air-water adiabatic | ||
Govan et al., 1989 [ | LOTUS test rig tubular section
| Air-water adiabatic | ||
Cousins and Hewitt, 1968 [ | acrylic resin tube | Air-water adiabatic | ||
Adamsson and Anglart, 2005
[ | Tube test section | Steam-water diabatic | ||
Okawa et al., 2005 [ | Deposition
mass transfer coefficients
Droplet
concentration
| Stainless steel tube
| Air-water | |
Fore et al., 2002 [ | Drop size distribution | Stainless steel duct | Nitrogen-water | Pressure 3.4 and 17 bar
Temperature |
Fore and Dukler, 1995 [ | Drop size distribution | Vertical
tube
| Air-water air-water+glycerine (50% mix) (6 cP liquid) | |
Andreussi,
1983 [ | Plexiglass tube | Air-water adiabatic | Pressure: atmospheric |
The following state of the art on the modeling of dryout by two-phase CFD results from the ongoing work in NURESIM.
(1)
(2)
(3)
(4)
(5)
While current industrial methods for CHF still use the subchannel analysis and empirical CHF correlations, the use of CFD already proved its potential interest in fine-scale investigations of boiling flows for a better understanding of sensitive flow processes. The “local predictive approach” where CHF empirical correlations would be based on local T/H parameters provided by CFD is not yet available but, with the present state of the modeling, CFD can already be used to subsequently help new fuel assembly design and to develop better CHF predictions in both PWR and BWR.
The NURESIM project is partly funded by the European Commission within the Sixth Framework Programme.