Film flow measurements at several axial positions in round pipes with variousaxial power distributions are presented for conditions corresponding to normaloperation of a BWR. It is confirmed that the film flow rate approaches zero atthe onset of dryout. Selected phenomenological models of annular two-phaseflow are shown to reasonably reproduce the measurements. It is concluded thatmodels are in better agreement with measurements if terms corresponding topossible boiling induced entrainment are excluded.

A method to perform film flow analysis in subchannels as a post-processto a standard two-field subchannel code is suggested. It is shown that thisapproach may yield accurate prediction of dryout power in rod bundles to alow computational cost and that the influence of the power distribution is wellpredicted by the model.

Measurements of film mass flow rate for annular, diabatic steam-water flow in tubes are presented. The measurements were carried out with four axial power distributions and at several axial positions at conditions typical for boiling water reactors, i.e. 7 MPa pressure and total mass flux in a range from 750 to 1750 kg/m2s. The results show that the influence of the axial power distribution on the dryout power corresponds to a consistent tendency in the film flow rate and that the film tends to zero when dryout is approached. Furthermore it is demonstrated that two selected phenomenological models of annular flow well predict the present data. A model for additional entrainment due to boiling is shown to degrade the predictions.

Measurements of developing films in adiabatic high pressure steam-water flow in annular geometry have been reanalyzed and compared to a linearized film-flow model. The development rate of the outer film could be determined with good accuracy in four cases. In one case it was also possible to conclude that the inner film develops faster than the outer one. When compared to the linearized model, these observations show that the deposition rate has to be almost independent of the drop concentration at the investigated conditions. Furthermore, any significant deposition by direct impaction of drops can be excluded as it would couple the development of the two films. These conclusions are quite general and do not depend on the use of any particular correlation for the deposition or entrainment rates. Finally, a rough estimate of the deposition rate was possible, confirming that deposition rates are considerably lower at high pressure steam-water flows than in air-water flows.

The motion of liquid drops in annular two-phase flow in pipes, annuli and subchannels has been investigated with a model based on Lagrangian particle tracking. The results confirm that large drops may deposit by direct impaction. It is also demonstrated that the deposition rate does not differ significantly between pipes and subchannels except for very large drops, which deposit slower in subchannels. Furthermore the Saffman lift force is shown to have a large impact on the results but it is questionable of the standard formulation is applicable to the drops considered here. Finally it is concluded that accurate modeling of high pressure steam-water flows requires a model for drop-drop collisions.

This paper was published when the measurements with non-uniform powerdistribution were still ongoing. Therefore only the measurements with uniformand top-peaked power profiles were included. The paper compares the measured data with deposition and entrainmentmodels by Hewitt & Govan (1990) and Okawa et al. (2003). These models are also discussed in Sections 4.2.1 and 4.2.2. The issue of a correct boundarycondition at the onset of annular flow was avoided by starting the integrationof the film flow model from the most upstream measurement point. In this way the net mass exchange rate (deposition less entrainment) could be studied without any initial bias from the boundary condition.The entrainment correlation proposed by Okawa et al. (2003) included a heat flux dependent term to account for boiling entrainment (Section 4.2.2). Paper 2 concludes that the model agrees better with measurements if this term is omitted. The result suggests that boiling entrainment may not be an important effect at the investigated conditions.

Measurements of film flow rates in diabatic annular flow in tubes with various axial power distributions were carried out in the high-pressure two-phase flow loop at the Royal Institute of Technology (KTH), Sweden. The measurements were performed at conditions typical for boiling water reactors, i.e. 7 MPa pressure and total mass flux in a range from 750 to 1750 kg/m(2)s. Four different axial power distributions were used and the film mass flow was measured at 7 axial locations for each set of boundary conditions. The results show that the outlet peaked distribution gives less film than the inlet peaked one. This result is consistent with well known trends from measurements of dryout power. The measurements also show that the film flow at the onset of dryout is very small at investigated conditions in agreement with earlier studies. Finally it is shown that the present data is well predicted by two selected phenomenological models of annular flow.

The influence of axial power distributions on dryout occurrence in nuclear fuel assemblies has been studied extensively for several decades. Even though it is well known that axial power shapes which may significantly vary in nuclear reactors during their operation significantly change the dryout power level, this particular influence is rather difficult to capture with current correlations. In this paper it is shown that this influence can be captured using a phenomenological liquid film model coupled to a standard sub-channel code. The model has been applied to various geometries, including a round pipe, as well as 5 x 5 and 8 x 8 fuel rod assemblies, and highly accurate predictions have been obtained.

This paper presents film flow measurement technique and the results with uniform power distribution. Based on these measurements it is possible to estimate the critical film thickness. The measured film thickness was plotted versus steam quality and slightlyextrapolated up to the measured critical steam quality. The conclusions werein line with Hewitt et al. (1965), i.e. that the critical film thickness is insignificantly small. This does not contradict e.g. Ueda & Isayama (1981) since the conditions were not the same, but for the flow conditions and heat fluxes that are typical for BWR operation it was concluded that the critical film thickness is, for practical purposes, zero.

An experimental study on post dryout heat transfer regime in annuli with flow obstacles wasconducted in the High-pressure Water Test (HWAT) loop at the Royal Institute of Technologyin Stockholm, Sweden. An annulus with flow obstacles, consisting of two concentric heatedpipes (12.7x24.3) mm, with total heated length equal to 3650 mm was employed as a testsection. The experimental investigations were performed in a wide range of the operationalconditions: mass flux (500-1750) kg/(m2s), inlet subcooling (10-40) K and system pressure(5-7) MPa. The wall superheat was measured at 88 different axial positions. A significanteffect of the flow obstacles on the wall temperature has been observed. A new correlation toevaluate the wall superheat in the post-dryout developing region and downstream of the flowobstacles was suggested. The new approach is taking into account in a combined manner theonset of the dryout point and the flow obstacle location. The coefficients and constants of thecorrelation have been optimized based on 1211 points obtained experimentally. Thecorrelation is applicable starting with the point of the onset of the dryout towards fullydeveloped post-dryout heat transfer regime and shows a correct asymptotical trend. Toaccount for the flow obstacle effect on the critical quality, an expression similar to theLevitan-Lanstman dryout correlation is suggested. The newly developed methodology can beused to predict the wall temperature in the post-dryout heat transfer regime over a wide rangeof mass fluxes and pressures typical for boiling water reactors.

The influence of flow obstacles on post-dryout heat transfer at typical BWR operational conditions has been investigated in bilaterally heated annuli. The objective of the study is two-fold: (a) capture the net effect of various obstacles by comparing the experimental results obtained in the “obstacle-free” test section with the results obtained in a test section with obstacle; (b) obtain a high spatial resolution of wall temperature measurements to allow for a precise determination of the dry-patch location in the heated channel. Present study provides an extensive experimental database with more than 22000 heat transfer coefficient values in pre-, trans- and post-dryout regimes in a wide range of operational conditions: pressure equal to 5 and 7 MPa, water mass flux from 500 to 1750 kg/m2.s, inlet subcooling equal to 10 and 40 K. The heated wall temperature has been measured with 88 thermocouples located along 3.65 m long annulus with diameters 12.7x24.3 mm.

New experimental studies of the onset of dryout and post-dryout heat transfer have been performed in the High-pressure Water Test (HWAT) loop at the Nuclear Reactor Engineering division, KTH, Stockholm, Sweden. The experiments have been performed in a bilaterally heated annulus with dimensions 12.7x24.3x3650 mm and with various flow obstacles placed in the exit part of the channel. The objective of the study has been to obtain a new data set of high accuracy which can be used for validation of detailed computational models for prediction of the influence of flow obstacles on the occurrence of dryout and on the post-dryout heat transfer. To meet the objective, in total 88 K-type thermocouples have been installed in the test section, providing both lateral and axial distribution of the heated wall temperature. Several thermocouples have been placed within and in a direct vicinity of selected flow obstacles to obtain a high spatial resolution of the measured temperature field. A thorough analysis of the experimental uncertainties indicates that the accuracy of temperature measurements is better than +/-2 K.

The measurements have been performed at conditions relevant to nuclear reactor safety applications: system pressure in a range from 5 to 9 MPa, mass flux from 500 to 1500 kg/(m².s) and inlet sub-cooling from 10 to 40 K. The heat flux applied in the test section was limited to not allow the wall temperature to exceed 900 K.

The experimental results indicate that flow obstacles can either remove a dry-patch completely, or reduce the wall temperature downstream of their location. It has been noted that this effect depends on operational conditions and also on the geometry and axial location of the obstacle. In general, stronger influence has been noted for high mass flow rates and for obstacles with a larger projected cross-section area.

An experimental study on post dryout heat transfer regime in annuli with flow obstacles was conducted in the High-pressure Water Test (HWAT) loop at the Royal Institute of Technology in Stockholm, Sweden. The experimental investigations were performed in a wide range of the operational conditions: mass flux (500-750) kg/(m²s), inlet subcooling (10-40) K and system pressure (5-7) MPa. The wall superheat was measured at 88 different axial and azimuthal positions (40 on the inner tube and 48 on the outer tube). The results show an enhancement of heat transfer downstream of flow obstacles. The biggest influence has been observed in case of pin spacers. This result is consistent with blockage area of various obstacles, which was the highest in case of pin spacers

This paper describes a new approach to predict post-dryout heat transfer in channels with flow obstacles. Using experimental data obtained in annular test sections at prototypical BWR conditions, the existing Saha correlation for post-dryout heat transfer has been modified to account for the presence of obstacles. The obstacle effect is taken into account in the following way: (a) the critical quality downstream of an obstacle is found; (b) an exponential function of equilibrium quality is applied to describe the decrease of heat transfer coefficient in the developing post-dryout region; (c) an additional heat transfer enhancement term is applied downstream of the obstacle. The new approach is applied to annular test sections with various flow obstacles and a significant improvement of accuracy of wall temperature prediction, as compared to reference methods, is shown.

The present paper presents results of the TRACE code assessment against post dryout experimental data obtained in tubes and annuli. The investigations have been focused on the experiments carried out at 70 bar pressure, 10 K inlet subcooling and the mass flux variation between 500-2000 kg/(m²s). Various axial power distributions (uniform, inlet peaked, middle peaked, and outlet peaked) have been used in the tube geometry. Uniform power distributions, with various ratios between the inner and the outer power have been used in the annulus geometry. The validation results indicated that the discrepancies between measured and calculated data are in a range of ± 20%. In addition a sensitivity study has been performed in which the influence of several parameters (including cell size and boundary conditions) on computational results have been investigated.

In this paper a new approach to predict wall temperature during post-dryout heat transfer in annuli with flow obstacles is presented. The proposed approach takes into account the obstacle specifics and location in the channel to determine the onset of post-dryout patch. The wall temperature in the dry patch area is predicted from a correlation that is taking into account the developing post-dryout heat transfer regime. The method is applied to post-dryout conditions in annulus with pin-spacers and a significant improvement of prediction accuracy in comparison to other reference methods is demonstrated.

An experimental study on post-dryout heat transfer was conducted in the High-pressure WAter Test (HWAT) loop at the Royal Institute of Technology in Stockholm, Sweden. The objective of the experiments was to investigate the influence of flow obstacles on the post-dryout heat transfer. The investigated operational conditions include mass flux equal to 500 kg/m² s, inlet sub-cooling 10 K and system pressure 5 and 7 MPa. The experiments were performed in annuli in which the central rod was supported with five pin spacers. Two additional types of flow obstacles were placed in the exit part of the test section: a cylinder supported on the central rod only and a typical BWR grid spacer cell. The measurements indicate that flow obstacles improve heat transfer in the boiling channel. It has been observed that the dryout power is higher when additional obstacles are present. In addition the wall temperature in post-dryout heat transfer regime is reduced due to increased turbulence and drop deposition. The present data can be used for validation of computational models of post-dryout heat transfer in channels with flow obstacles.

An experimental study on post dryout heat transfer regime in annuli with flow obstacles was conducted in the High-pressure Water Test (HWAT) loop at the Royal Institute of Technology in Stockholm, Sweden. An annulus consisting of two concentric heated pipes (12.7x24.3) mm, with total heated length equal to 3650 mm was employed as a test section. Three kinds of flow obstacles were used: pin-spacers, cylindrical obstacles and grid obstacles. The experiments performed in the test section with pin-spacers only were considered as the reference case. In two consecutive sets of runs, additional obstacles were placed inside the flow channel while keeping the pin spacers in the same positions. In that way the net effect of obstacles on heat transfer was measured. The experimental investigations were performed in a wide range of the operational conditions: mass flux (500-1750) kg/(m²s), inlet subcooling (10-40) K and system pressure (5-7) MPa. The wall superheat was measured at 88 different axial positions (40 on the inner tube and 48 on the outer tube) for the conditions mentioned above. A local heat transfer coefficient was calculated based on the measured annulus wall temperatures and the saturated fluid (water) properties. The results show an enhancement of the heat transfer coefficient downstream of flow obstacles. The most significant influence has been observed in case of pin spacers. This result is consistent with blockage area of various obstacles, which was the highest in case of pin spacers. The data obtained in more than 200 runs were compared with two pre-dryout and post-dryout correlations. The correlations show a slight over-prediction of the heat transfer coefficient in both pre-dryout and post-dryout regions. The thesis contains a detailed description of experimental procedures as well as an analysis of the results of measurements.

New experiments have been performed to investigate heat transfer to water/steam two-phase mixture flowing in annular test section at trans-dryout conditions. The measurements have been carried out in the High-pressure Water Test (HWAT) loop at the Royal Institute of Technology, Stockholm, Sweden. The primary objective of the experimental investigations has been to study heat transfer at conditions typical for Boiling Water Reactors (BWR), when heat transfer regime changes from convective boiling to post-dryout heat transfer. The experiments indicate a significant enhancement of heat transfer just upstream of dryout patch. It has been observed that the measured heat transfer coefficient is in good agreement with the Chen correlation for quality less than 30%, however, increasing discrepancy is noted for near-critical quality.

Post-dryout heat transfer to high pressure water was investigated experimentally in vertical tubes and annuli containing various flow obstacles. The operational conditions during the experiments were as follows: mass flux from 500 to 1750 kg/m(2) s. pressure from 5 to 9 MPa, inlet subcooling from 10 to 40K and heat flux up to 1.5 MW/m(2). Five different test sections were used in experiments: three annular test sections with inner diameter 12.7 mm and outer diameter 24.3 mm, containing cylindrical and grid flow obstacles in the upper part, and two tubular test sections with inner diameter 24.3 mm with and without pin flow obstacles. The heated length in all test sections was 3650 mm. The wall temperature was measured with 88 thermocouples located along the inner rod and the outer tube surfaces. Due to the presence of flow obstacles, only developing post-dryout heat transfer was observed. Selected post-dryout heat transfer correlations were compared to the experimental data. It has been concluded that all tested correlations predict significantly higher wall temperatures than those obtained in the present experiment. A simple correction function to the Saha model has been suggested which significantly improves the agreement between the correlation and the present data.

An experimental study on post-dryout heat transfer was conducted in the High-pressure WAter Test (HWAT) loop at the Royal Institute of Technology in Stockholm, Sweden. The objective of the experiments was to investigate the influence of flow obstacles on the post-dryout heat transfer. The investigated operational conditions include mass flux equal to 500 kg/(m²s), inlet sub-cooling 10 K and system pressure 5 and 7 MPa. The experiments were performed in annuli in which the central rod was supported with five pin spacers. Two additional types of flow obstacles were placed in the exit part of the test section: a cylinder supported on the central rod only and a typical BWR grid spacer cell. The measurements indicate that flow obstacles improve heat transfer in the boiling channel. It has been observed that the dryout power is higher when additional obstacles are present. In addition the wall temperature in post-dryout heat transfer regime is reduced due to increased turbulence and drop deposition. The present data can be used for validation of computational models of postdryout heat transfer in channels with flow obstacles.

The purpose of this paper is to present the experimental setup, experimental method and results of the recent post-dryout heat transfer investigations in an annulus with pin spacers. The experiments were performed in the thermal-hydraulic laboratory at the Royal Institute of Technology (KTH), Stockholm, Sweden. The experimental facility has an annular test section which consists of two double-heated concentric tubes manufactured of Inconel 600. Five levels of pin spacers were installed along the test section to keep the rod and the tube equidistant during experiments. The experimental matrix includes measurements of wall temperature distributions for single phase and twophase flows, for both convective boiling and postdryout heat transfer regimes. The influence of variations in mass flux (500-1500) kg/(m²s) and inlet subcooling 10 and 40 K at system pressure of 7 Mpa were investigated. The experimental results indicate that post dryout heat transfer is influenced by the pin spacers. In particular it has been observed that the dry patch appearing in the test section can be quenched downstream of the pins-spacer. The current results provide additional  experimental database which can be used for validation of post-dryout heat transfer models including the flow obstacle effects.

This paper describes the experimental setup, instrumentation and procedures which have been developed in the thermal-hydraulic laboratory at the Royal Institute of Technology (KTH), Stockholm, Sweden, to perform new post-dryout heat transfer investigations in an annulus with flow obstacles. Previous investigations performed in the same laboratory indicated that flow obstacles had a considerable influence on the post-CHF heat transfer. The measured heat transfer enhancement was significantly under-predicted by existing models. However, the net effect of obstacles could not be deduced from the measurements, since reference - obstacle-free measurements- had not been performed. In addition, the number of thermocouples that could be installed inside the heated rod was limited to 8. These deficiencies have been removed in the current approach. Firstly, the present design of the test section allows for measurements both with and without flow obstacles. In this way the net effect of the obstacles will be captured. Secondly, a newly developed technique allowed the installation of 40 thermocouples inside of the heated rod. An additional 40 thermocouples have been installed on the external wall of the heated tube. Therefore, a significant improvement of the accuracy of measurements can be expected. The present arrangement of instrumentation is suitable to perform measurements of heat transfer under both steady-state and transient conditions.

The development of nuclear industry is influenced by the following four major factors: plant safety, non-proliferation of nuclear materials, economics and sustainability. Recently, the safety records of nuclear power plants are excellent and it can be believed that this factor will not limit the development in the nuclear field in the future. The same can be said about the non-proliferation, even though care must be taken to limit the number of countries that will try to develop all facilities that are needed for fuel cycle operations. It is believed that the future development of nuclear power will be – to a large extend – governed by over-all economics of this type of electricity generation. This paper gives an overview of the major aspects of an economic analysis of a nuclear power plant. The method of calculation of the levelised cost of electricity is described. Selected data found in literature indicate that nuclear electricity production will be economically favorable when CO₂ emission charge is taken into account for the coal and gas electricity production. Additional improvement of nuclear economy will be obtained by the shortening of the construction period, and in general, by reduction of the capital cost during the construction of a plant.

In this paper analytical solutions have been obtained for the case of laminar supercritical water flow between two infinite walls at two slightly different temperatures. The existence of the fully-developed flow condition is postulated and thus the problem is reduced to solutions of a set of non-linear ordinary differential equations representing the conservation of mass, energy and momentum. The equations are numerically solved using the Matlab code, and employing water property functions based on the IAPWS-IF97 library. The Matlab program has been verified against known analytical solutions for the laminar duct flow of fluids with constant properties and an excellent agreement has been obtained. The solutions of conservation equations give an expression for the over-all heat transfer coefficient in mixed convection. The analysis indicates that the heat transfer coefficient is a very strong function of local properties and flow conditions. The predicted Nusselt number and the convection coefficient are in a reasonably good agreement with existing correlations once the fluid temperature is far from the pseudo-critical value. However, the existing correlations deviate from the predicted values when the fluid is at the pseudo-critical temperature.

An analysis is presented that predicts the conditions which allow for a formation of a stable dry patch in diabatic annular two phase flows. The analysis employs a force balance formulated for the leading edge of the liquid film In addition to stagnation, thermo-capillary and vapor thrust forces, the analysis includes effects of the pressure gradient and the interfacial shear stress. It is shown that the equilibrium conditions of a dry patch are dominated by the stagnation force, the surface tension force, the capillary force and the skin drag force. For high heat flux conditions only the first three forces are important.

One of the important parameters that affect the thermal-hydraulic performance of nuclear fuel assemblies is the spatial - that is the lateral and the axial distribution of power. Since this parameter may have a significant influence on thermal margins of nuclear reactors, it is necessary to take it into account in various models and/or correlations. One practical difficulty in doing so is the fact that the spatial power distribution is a function of space variables, which makes it very inconvenient to implement into single-parameter correlations. In addition, there is still lack of a simple theoretical model that captures the effect of spatial power distributions on the thermal-hydraulic performance of nuclear fuel assemblies. In this paper, an accurate and fast running convolution method is presented to predict the influence of axial power distribution on wall temperature distributions. The method has been verified against CFD predictions of the wall temperature in a heated pipe and an excellent agreement between the two approaches is demonstrated. The method is applicable only for constant fluid properties and for a fully developed flow regime.

Supercritical water will serve as a coolant in the Generation-IV Supercritical Water-Cooled Reactor (SCWR). The important advantage of supercritical water as a coolant is the lack of the phase-change phenomenon. As a result one of the most limiting factors applicable to the current Light Water Reactors (LWR) – namely the occurrence of the Critical Heat Flux (CHF) – is no longer existent. Considering the high heat capacity, supercritical water is indeed an excellent choice for a coolant. However, even though CHF is no longer an issue, heat transfer to supercritical water suffers from a sudden Heat Transfer Deterioration (HTD) phenomenon. HTD manifests itself with a sudden reduction of the local heat transfer coefficient and local increase of the heater wall temperature. Even though the phenomenon has been intensively investigated in the past 50 years, there is a lack of a robust and accurate criterion for the onset of HTD.

Recently, Palko and Anglart (2007) demonstrated that the onset of HTD can be captured with a computational model based on the Reynolds Averaged Navier Stokes (RANS) equations and using the Shear-Stress Transport (SST) turbulence model implemented in the CFX code (Menter, 1993).  The calculations revealed that there are two principal mechanisms of the onset of HTD: (a) reduction of the turbulence intensity close to the wall due to the buoyancy effects, (b) creation of a thin layer of supercritical water with low thermal conductivity (corresponding to the “vapor” phase of the supercritical fluid). The former mechanism occurs for relatively low mass fluxes, whereas the latter occurs when the mass flux of coolant is pretty high. The present paper presents further numerical investigation of the HTD phenomenon, and in particular, a derivation of the criterion for the onset of HTD based on numerical simulations.

In this paper the mechanism of the onset of heat transfer deterioration to supercritical water is elucidated with detailed numerical predictions of flow and heat transfer in the boundary layer. It is shown that for low mass flow rates the buoyancy effects are dominant and the deterioration of heat transfer is caused by the turbulence damping in the vicinity of the heated wall. For high mass flow rates the mechanism of deterioration changes and the triggering factor is the decrease of the thermal conductivity of fluid in the viscous sub-layer. A numerical prediction of this phenomenon requires application of a low Reynolds number turbulence model with y⁺ less than 1.

The recent focus on the global warming caused by greenhouse gases has drawn attention to thenuclear energy as a reliable, carbon-dioxide free source of electricity. The current optimistic view onthe nuclear energy can be compared to the situation that took place in sixties and early seventies ofthe past century, when the potential of the nuclear energy seemed to be almost unlimited. However,today’s enthusiasm, frequently referred to as the nuclear renaissance, is different. The nuclearcommunity learned its lessons, such as resulting from the two major accidents in Three Mile Islandand in Chernobyl, but also resulting from the competition with other energy sources and – last butnot least – due to the increasing public awareness of nuclear safety issues. Thus, the current trends ofdevelopment in nuclear energy engineering have different priorities than in the past. Currentlydesigned and constructed nuclear power plants will have so-called generation III+ reactors. Suchreactors are designed with focus on safety and efficiency aspects. Future generation IV reactors arecurrently under development and subject of intensive world-wide research. These reactors will notonly be safe and efficient, but also resistant to proliferation and will support the sustainabledevelopment in the nuclear energy field.

Conditions for the formation of a stable dry patch in vertical annular two-phase flows in heated channels are investigated. An analytical model of the force balance for the leading edge of the liquid film is developed. In addition to surface tension, evaporation thrust and capillary forces, the model includes the effect of turbulence, the pressure gradient and the interfacial shear stress. Numerical evaluations are performed to validate the model and to indicate the importance of various factors on the dry patch stability and on the resulting minimum wetting rate of the liquid film. The analyses indicate that good agreement with measurements is obtained in case of stagnant patch formed on liquid film flowing down a vertical surface. It is shown that for low and moderate mass flow rate of the gas phase in vertical co-current annular flow, the force balance is dominated by the stagnation and the shear stress forces. With growing mass flow rate of the gas phase, the pressure gradient and the interfacial shear stress are increasingly important. As a result, in accordance with measurements, the predicted minimum flow rate of the liquid film at which the patch is re-wetted decreases.

One of the major objectives of HPLWR Phase 2 project is to assess the feasibility ofthe super-critical water-cooled and moderated reactor. To this end it is necessary toevaluate the thermal performance and the distribution of temperature in the reactorcore under target operation conditions (25 MPa pressure and 280÷500 °C temperature).The critical scientific issue in this subject is to evaluate the efficiency of heattransfer in the reactor core and in particular, to determine the conditions of the occurrenceof the Heat Transfer Deterioration (HTD) phenomenon. The objective of this paperis to investigate and rank the key parameters that are governing the onset of HTDin general and in the HPLWR in particular. The paper describes the major heat transferdeterioration mechanisms and criteria, and contains evaluation of the importance ofparticular mechanisms on the onset of HTD. The major mechanisms are listed in a tableand ranked according to their estimated relative importance. It is concluded thateven though HTD has been investigated by many researchers and the world-wide experimentaldatabase counts tens of thousands of experimental points; still prediction ofthe onset of HTD is difficult and is subject to essential uncertainties. As a result, thereare many expressions for the onset of HTD available; however, they give very oftennon-consistent predictions. Thus more work is needed to develop a reliable HTD criterionthat can be used in nuclear applications. Operational conditions and complex geometryof HPLWR introduce additional uncertainties into the prediction of the onset ofHTD. Limited experiments in rod bundles and in parallel channels indicate a significantdeparture of the HTD conditions as compared to the experiments in single channels.Such factors as the bundle effect and the flow oscillation effect (in case of parallelchannels) have been noted and discussed.

Post-dryout heat transfer plays an important role in safe and economical operations of Light Water Reactors (LWR). This type of heat transfer is avoided under normal operational conditions of nuclear reactors; however, it may occur in transient or accidential situations. To estimate the risk of clad damages due to increase of temperature associated with the occurrence of post-dryout, it is necessary to properly model heat transfer processes under such conditions. The influence of various parameters on heat transfer downstream of spacer has been investigated. It is concluded that heat transfer enhancement due to spacers is largely under-predicted for flows with relatively low quality. For such flows the effect of droplets impinging heated walls is significant and must properly be taken into account. The phenomenological model presented in this paper shows a superior accuracy over correlations and presents a potential to capture the phenomenon of rewetting  that occurs downstream of spacers.

Post-dryout heat transfer plays an important role in safe and economical operations of Light Water Reactors (LWR). This type of heat transfer is avoided under normal operational conditions of nuclear reactors; however, it may occur in transient or accidential situations. To estimate the risk of clad damages due to increase of temperature associated with the occurrence of post-dryout, it is necessary to properly model heat transfer processes under such conditions. The influence of various parameters on heat transfer downstream of spacer has been investigated. It is concluded that heat transfer enhancement due to spacers is largely under-predicted for flows with relatively low quality. For such flows the effect of droplets impinging heated walls is significant and must properly be taken into account. The phenomenological model presented in this paper shows a superior accuracy over correlations and presents a potential to capture the phenomenon of rewetting that occurs downstream of spacers.

One of the challenges in thermal-hydraulic analyses of BWRs is correct prediction of dryout occurrence in fuel assemblies. In practical applications the critical powers in fuel assemblies are found from correlations that are based on experimental data. The drawback of this approach is that correlations are valid only for these fuel assemblies on which the experiments have been conducted. Other restrictive factors are the limited ranges of experimental working conditions including pressure, mass flux and axial power distributions. To overcome the above-mentioned limitations, several different approaches have been proposed to predict the dryout occurrence. One of them is to employ a phenomenological model of annular flow, in which the mass transfer between the liquid film and the gas core is based on entrainment and deposition correlations. Most of these correlations are derived from water-air flows in vertical tubes and their applicability to other geometries in general, and rod-bundles in particular, should be analysed. This paper presents an analysis of the entrainment rate in vertical annuli. Using the standard approach to calculate the entrainment rate, one can demonstrate that the results deviate from measurements. It has been shown that modifying the entrainment correlation based on data obtained in the annulus geometry leads to an essential improvement in the predictive capability of the phenomenological model of annular two-phase flow.

In this paper the mechanism of the onset of heat transfer deterioration to supercritical water is elucidated with detailed numerical predictions of flow and heat transfer in the boundary layer. It is shown that for low mass flow rates the buoyancy effects are dominant and the deterioration of heat transfer is caused by the turbulence damping in the vicinity of the heated wall. For high mass flow rates the mechanism of deterioration changes and the triggering factor is the decrease of the thermal conductivity of fluid in the viscous sub-layer. A numerical prediction of this phenomenon requires application of a low Reynolds number turbulence model with y+ less than 1.

Conditions for the formation of a stable dry patch in vertical diabatic annular two-phase flows are investigated. An analytical model of the force balance for the leading edge of the liquid film is developed. In addition to surface tension, evaporation thrust and capillary forces, the model includes the effect of turbulence, the pressure gradient and the interfacial shear stress. Numerical evaluations are performed to validate the model and to indicate the importance of various factors on the dry patch stability and on the resulting minimum wetting rate of the liquid film. The analyses indicate that good agreement with measurements is obtained in case of stagnant patch formed on liquid film flowing down a vertical surface.

This paper presents principles of passive and active cooling that are suitable to mirror-based hybrid, nuclear fission/fusion systems. It is shown that liquid metal lead-bismuth cooling of the mirror machine with 25 m height and 1.5 GW thermal power is feasible both in the active mode during the normal operation and in the passive mode after the reactor shutdown. In the active mode the achievable required pumping power can well be below 50 MW, whereas the passive mode provides enough coolant flow to keep the clad temperature below the damage limits.

Axial power distribution is one of the parameters that influence the occurrence of the dryout in nuclear fuel assemblies. Experimental data indicate that this influence is quite substantial, ranging from few to above ten percent of the total power. Thus accurate prediction of the dryout power for various power distributions has important implications on the economy and safety of nuclear power plants. The difficulty with capturing the influence of that parameter stems from the fact that during reactor operation practically unlimited number of power shapes can occur. This fact makes it very difficult to investigate the effect experimentally, and an analytical approach is needed. Various methods have been proposed in the past to capture the effect of non-uniform power distribution on dryout. These approaches can be divided into several categories, where the two main ones are as follows: (a) methods based on introduction of a shape factor, which is calculated from the known shape of the power distribution; (b) methods using certain integral parameters, such as the boiling length and the annular flow length, which are expressed as functions of axial power distribution. In the present approach a simplified annular flow model is used, in which the dryout occurrence is based on the prediction of the disappearance of the liquid film. The dependence of the dryout power on the axial power shape is obtained in a general analytical form. Based on this analytical solution, a new set of terms that govern the dryout power in channels with various axial power distributions is proposed.

The present paper deals with a theoretical analysis of the spray cooling of a Reactor Pressure Vessel (RPV) head in a Boiling Water Reactor (BWR). To this end a detailed computational model has been developed. The model predicts the trajectories, diameters and temperatures of subcooled droplets moving in saturated vapor. The model has been validated through comparison with experimental data, in which droplet temperatures were measured as functions of the distance that they cover in saturated vapor from the moment they leave the sprinkler outlet to the moment they impact on the RPV head inner wall. The calculations are in very good agreement with measurements, confirming the model adequacy for the present study. The model has been used for a parametric study to investigate the influence of several parameters on the cooling efficiency of the spray system. Based on the study it has been shown that one of the main parameters that govern the temperature increase in a subcooled droplet is its initial diameter. Comparisons are also made between conclusions from the theoretical model and observations made through flow and temperature measurements in the plant (Forsmark 1 and 2). One of these observations is that the rate at which the RPV head temperature decreases on the way down from hot to cold standby is constant and independent of the sprinkling flow rate as long as the flow rate is above a certain minimum value. Accordingly, the theoretical model shows that if one assumes that the cooling of the RPV head is through a water film built on the inner wall due to sprinkling, the heat removal rate is only very weakly dependent on the sprinkling flow rate.

The present paper deals with a theoretical analysis of the spray cooling of a Reactor Pressure Vessel (RPV) head in BWR. To this end a detailed computational model has been developed. The model predicts the trajectories, diameters and temperatures of subcooled droplets moving in saturated vapor. The model has been validated through comparison with experimental data, in which droplet temperatures were measured as functions of the distance that they cover in saturated vapor from the moment they leave the sprinkler outlet to the moment they impact on the RPV head inner wall. The calculations are in very good agreement with measurements, confirming the model adequacy for the present study. The model has been used for a parametric study to investigate the influence of several parameters on the cooling efficiency of the spray system. Based on the study it has been shown that one of the main parameters that govern the temperature increase in a subcooled droplet is its initial diameter. Comparisons are also made between conclusions from the theoretical model and observations made through flow and temperature measurements in the plant (Forsmark 1 and 2). One of these observations is that the rate at which the RPV head temperature decreases on the way down from hot to cold standby is constant and independent of the sprinkling flow rate as long as the flow rate is above a certain minimum value. Accordingly, the theoretical model shows that if one assumes that the cooling of the RPV head is through a water film built on the inner wall due to sprinkling, the heat removal rate is only very weakly dependent on the sprinkling flow rate.