Ladislav Vyskočil - Academia.edu (original) (raw)
Papers by Ladislav Vyskočil
Science and Technology of Nuclear Installations, 2009
The NURESIM Project of the 6th European Framework Program initiated the development of a new-gene... more 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 applic...
<jats:p>The paper describes methods used for evaluation of pressurized thermal shock (PTS) ... more <jats:p>The paper describes methods used for evaluation of pressurized thermal shock (PTS) at UJV Rez and selected results. The UJV Rez participated on the one hand in a number of international programs focused on development of PTS methodology, benchmark testing etc. On the other hand UJV has worked in several projects evaluating PTS for nuclear power plants in the Czech Republic and abroad. The introductory part of the paper also briefly mentions a development of PTS methodology in the world, its current status and trends.</jats:p> <jats:p>The core of the paper describes thermal hydraulic part of the PTS evaluation process, which usually contains system thermal hydraulic (TH) analysis and mixing calculation. Already at the system TH calculation level, a simple 2D model of reactor downcomer (DC) is used in UJV Rez analyses — as this approach enables the prediction of 2D temperature and velocity fields in DC and improves the prediction of natural circulation flow or its stagnation in individual loops. For the subsequent mixing calculation the Computational Fluid Dynamics (CFD) computer codes are used nowadays. Comparison of 2D temperature fields from 2D system TH analysis and CFD calculation shows surprisingly good agreement for most cases. The paper shows such comparison for several representative cases and takes a more detailed look into one of them.</jats:p>
Houille Blanche-revue Internationale De L Eau, 2009
Ce rapport decrit une simulation numerique d'experiences determinees a flux calorifique criti... more Ce rapport decrit une simulation numerique d'experiences determinees a flux calorifique critique effectuees sur une boucle experimentale "Large Water Loop ". Les calculs ont ete effectues a l'aide du programme NEPTUNE_CFD. "Large Water Loop" (LWL) est une boucle a pression hydraulique inactive dont les parametres technologiques et calorifiques sont appropries pour des reacteurs nucleaires a eau sous pression. Un dispositif experimental visant a rechercher un flux calorifique critique dans un faisceau de tiges electriquement chauffees fait partie du LWL. Les conditions d'une ebullition critique sont fixees a pression constante, a temperature constante de l'eau a l'entree et a debit constant, le rendement calorifique du faisceau augmentant peu a peu. Les tiges sont des tubes creux profiles a echauffement direct des parois. Les resultats des calculs de quatre experiences typiques sont presentees dans le rapport. On a pu noter a chaque fois un acc...
This paper presents an attempt to use multiphase CFD code for prediction of the Departure from Nu... more This paper presents an attempt to use multiphase CFD code for prediction of the Departure from Nucleate Boiling (DNB) type of Critical Heat Flux (CHF). Numerical simulations of DNB in boiling flow in vertical tube were performed with the NEPTUNE_CFD V2 code. This code can simulate multicomponent multiphase flow by solving three balance equations for each phase or fluid component. A new set of validated models of physical phenomena in boiling bubbly flow was used in the calculations. Simulated cases were based on data from the Standard tables of CHF in pipes, produced by the Russian Academy of Sciences. It was found out that local DNB criterion based on void fraction equal to 0.8 does not work well with the new set of physical models implemented in NEPTUNE_CFD V2 code. But it was discovered that the criterion for DNB prediction can be based on the ratio of evaporation heat flux and total wall heat flux. Evaporation heat flux is calculated by the extended Kurul and Podowski wall boili...
This paper presents numerical simulations of the boiling flow in a tube with a Departure from Nuc... more This paper presents numerical simulations of the boiling flow in a tube with a Departure from Nucleate Boiling type of critical heat flux (CHF). Standard tables of CHF produced by the Russian Academy of Sciences were used as a data set. The simulations were performed with the multiphase code NEPTUNE_CFD V1.0.7. A simple criterion based on the void fraction at the wall was used for the CHF prediction. Four data series were selected from the tables. In every series, one of the following parameters was variable: the local equilibrium quality, the mass flux, pressure and the tube diameter. The remaining three parameters were fixed. In every data point, a numerical simulation was performed so as to find out the interval of the wall heat fluxes at which the boiling crisis occurs. NEPTUNE was able to quite accurately predict CHF in cases with high mass fluxes and high pressures. On the other hand, in one low-mass-flux case, the CHF in the calculation occurred at a wall heat flux as low as ...
This paper presents a CFD simulation of selected "Large Water Loop" critical heat flux ... more This paper presents a CFD simulation of selected "Large Water Loop" critical heat flux experiments. Calculations were performed by NEPTUNE_CFD code. The Large Water Loop (LWL) is non-active pressurised-water equipment with technological and thermal parameters corresponding to those of PWR. The CHF experimental facility (a part of the Large Water Loop) has been designed for research into CHF in water flow through a bundle of electrically heated vertical rods. The critical conditions were determined under constant pressure, inlet water temperature and mass flux and for quasi steady-state -by gradually increasing the heat input. The rods are modelled by hollow tubes with direct heating of the wall. Thirteen CHF tests were calculated with NEPTUNE. In all calculated tests, a sudden rise of the wall temperature was observed. Simulations of cases with a higher mass flux were successful. Simulation of cases with a low mass flux indicates that the modelling approach might not be su...
This paper presents simulations of the convective boiling flow performed with NEPTUNE_CFD and FLU... more This paper presents simulations of the convective boiling flow performed with NEPTUNE_CFD and FLUENT codes. The DEBORA experiments carried out at CEA Grenoble were used as an experimental data set. In these experiments, freon R12 flows upwards inside a vertical pipe. Radial profiles of the flow variables are measured at the end of the heated section. Seven DEBORA cases were selected for simulation. NEPTUNE_CFD code was used without modifications because it contains all necessary models. In FLUENT, an important part of the models has been implemented by programming in User Defined Functions. The comparison of the radial profiles of void fraction, liquid temperature, gas velocity and mean bubble diameter at the end of the heated section shows that both codes can provide reasonable results in boiling conditions. The presented work was carried out within the 6 th Framework EC NURESIM project. NEPTUNE_CFD code is implemented in the NURESIM platform.
This paper presents coupled system of thermal hydraulic code Athlet, neutron kinetic code Dyn3D a... more This paper presents coupled system of thermal hydraulic code Athlet, neutron kinetic code Dyn3D and CFD code Fluent. The coupled system is intended for simulation of complex transients such as main-steam-line-break scenarios, which cannot be modeled separately first by the system and neutron kinetic code and then by CFD code, because of the feedbacks between the codes. The presented method is limited to scenarios with single-phase flow in CFD domain. In the first part of this paper, the coupling method is described. Explicit coupling of overlapped computational domains is used in this work. Exchange of the variables at the coupling interfaces is presented as well as the implementation of the coupled system on the Linux system. The master program that controls both Fluent and Athlet/Dyn3D codes was developed. The second part of the paper presents the demonstration simulation performed by the coupled system of Fluent and Athlet/Dyn3D. “Opening of Steam Dump to the Atmosphere” test car...
Multiphase Science and Technology
Predictions of local boiling flow processes leading to departure from nucleate boiling (DNB) cond... more Predictions of local boiling flow processes leading to departure from nucleate boiling (DNB) conditions are considered. The work was performed within the Nuclear Reactor Simulations project (NURESIM) within the Sixth European Framework program. This paper focuses on the Reynolds-averaged Navier-Stokes (RANS) approach as being the most reliable for simulation of realistic bubbly flows. New physical models developed within the NURESIM project are presented and tested on various single-channel boiling experiments, differing in geometry, working fluid, and operating conditions. The applicability of the model for boiling in fuel rod bundles under industrial conditions has been demonstrated.
La Houille Blanche, 2009
ABSTRACT This report presents a CFD simulation of selected "Large Water Loop" c... more ABSTRACT This report presents a CFD simulation of selected "Large Water Loop" critical heat flux experiments. Calculations were performed by NEPTUNE_CFD code. The Large Water Loop (LWL) is non-active pressurised-water equipment with technological and thermal parameters corresponding to those of pressurized water reactors. The CHF experimental facility (a part of the Large Water Loop) has been designed for research into CHF in water flow through a bundle of electrically heated vertical rods. The critical conditions were determined under constant pressure, inlet water temperature and mass flux and for quasi steady-state - by gradually increasing the heat input. The rods are modelled by hollow tubes with direct heating of the wall. Calculation results of four typical tests are presented in this paper. In all calculated tests, a sudden rise of the wall temperature was observed. Simulations of cases with a higher mass flux were successful. Simulation of cases with a low mass flux indicates that the modelling approach might not be suitable for lower mass fluxes. The results show that NEPTUNE has potential for predicting boiling flow up to CHF in the geometry of nuclear reactor fuel assembly. The presented work was carried out as part of the NURESIM project. NEPTUNE_CFD code is implemented in the NURESIM platform. Ce rapport décrit une simulation numérique d'expériences déterminées à flux calorifique ritique effectuées sur une boucle expérimentale "Large Water Loop". Les calculs ont été effectués à l'aide du programme NEPTUNE_CFD. "Large Water Loop" (LWL) est une boucle à pression hydraulique inactive dont les paramètres technologiques et calorifiques sont appropriés pour des réacteurs nucléaires à eau sous pression. Un dispositif expérimental visant à rechercher un flux calorifique critique dans un faisceau de tiges électriquement chauffées fait partie du LWL. Les conditions d'une ébullition critique sont fixées à pression constante, à temperature constante de l'eau à l'entrée et à débit constant, le rendement calorifique du faisceau augmentant peu à peu. Les tiges sont des tubes creux profilés à échauffement direct des parois. Les résultats des calculs de quatre expériences typiques sont présentées dans le rapport. On a pu noter à chaque fois un accroissement rapide de la température des parois. Les simulations des expériences à fort écoulement ont été couronnées de succès. Les calculs des expériences à faible écoulement ont montré que le procédé indiqué ne convenait pas pour des faibles écoulements. Les résultats montrent que le programme NEPTUNE présente un réel potentiel pour les calculs d'écoulement avec ébullition ou point critique d'ébullition dans la géométrie d'un caisson de chauffage d'un réacteur nucléaire. Cette étude a été réalisée dans le cadre du projet européen NURESIM. Le programme NEPTUNE_CFD fait partie de la plateforme NURESIM.
Annals of Nuclear Energy, 2014
ABSTRACT The prediction of condensation in the presence of noncondensable gases has received cont... more ABSTRACT The prediction of condensation in the presence of noncondensable gases has received continuing attention in the frame of the Severe Accident Research Network of Excellence, both in the first (2004-2008) and in the second (2009-2013) EC integrated projects. Among the different reasons for considering so relevant this basic phenomenon, coped with by classical treatments dated in the first decades of the last century, there is the interest for developing updated CFD models for reactor containment analysis, requiring validating at a different level the available modelling techniques. In the frame of SARNET, benchmarking activities were undertaken taking advantage of the work performed at different institutions in setting up and developing models for steam condensation in conditions of interest for nuclear reactor containment.
Nuclear Engineering and Design, 2015
ABSTRACT This paper presents a benchmark performed in the frame of the SARNET-2 EU project, deali... more ABSTRACT This paper presents a benchmark performed in the frame of the SARNET-2 EU project, dealing with momentum transfer between a real-scale PWR spray and the surrounding gas. It presents a description of the IRSN tests on the CALIST facility, the participating codes (8 contributions), code-experiment and code-to-code comparisons. It is found that droplet velocities are almost well calculated one meter below the spray nozzle, even if the spread of the spray is not recovered and the values of the entrained gas velocity vary up to 100% from one code to another. Concerning sensitivity analysis, several ‘simplifications’ have been made by the contributors, especially based on the boundary conditions applied at the location where droplets are injected. It is shown here that such simplifications influence droplet and entrained gas characteristics. The next step will be to translate these conclusions in terms of variables representative of interesting parameters for nuclear safety.
Recently, the safety analyses of VVER and PWR reactors have dealt with the possibility of reactiv... more Recently, the safety analyses of VVER and PWR reactors have dealt with the possibility of reactivity-induced accidents related to the penetration of a water slug with low boron concentration into the reactor core. Loop seals at the reactor coolant pump (RCP) suction are the most likely places for the formation of these slugs. The slug is formed in the loop when there is neither natural nor forced circulation. When the circulation is restored, the slug travels towards the reactor and causes an insertion of positive reactivity in the core. This report deals with a CFD simulation of the most dangerous event - the start-up of the first RCP. Only several seconds are needed for slug to reach the core and the operator has no time for corrective action. Mixing of slug on its way to the core can reduce the danger of core recriticality. The primary objective of this study was to find out whether the FLUENT 6 CFD code is capable of predicting the mixing in the cold leg, downcomer and lower ple...
Nuclear Engineering and Design, 2014
ABSTRACT This article presents a custom condensation model for commercial CFD code Fluent. The co... more ABSTRACT This article presents a custom condensation model for commercial CFD code Fluent. The condensation model was developed for the species transport model in Fluent code and it is suitable for both compressible and incompressible flow of air–steam mixture with additional non-condensable gases. The condensation model consists of two parts: condensation in volume and condensation on the wall. Condensation in volume is modeled by “return to saturation in constant time scale” method. Condensation on the wall is calculated from diffusion of steam through a layer of non-condensable gases near the wall. The performance of the condensation model was tested on the CONAN experiments. In these experiments, air–steam mixture flows downwards through a vertical channel with square cross section. One vertical wall of the channel is cooled and the steam condenses on it. The same model was then applied in simulation of PANDA Test 9bis experiment with condensation. In this test, two vessels connected with a pipe were filled with air; and steam was released into the first vessel. As the steam concentration increased in the vessels, the steam started condensing on the walls. The results of CFD simulations of both CONAN and PANDA experiments compared well with the measured data.
Nuclear Engineering and Design, 2007
The EU project FLOMIX-R was aimed at describing the mixing phenomena relevant for both safety ana... more The EU project FLOMIX-R was aimed at describing the mixing phenomena relevant for both safety analysis, particularly in steam line break and boron dilution scenarios, and mixing phenomena of interest for economical operation and the structural integrity.This report will focus on the computational fluid dynamics (CFD) code validation. Best practice guidelines (BPG) were applied in all CFD work when choosing
The NURESIM Project of the 6th European Framework Program initiated the development of a new-gene... more 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 applic...
Nuclear Engineering and Design, 2014
h i g h l i g h t s • Coupling interface between CFD code Fluent and system code Athlet was creat... more h i g h l i g h t s • Coupling interface between CFD code Fluent and system code Athlet was created. • Athlet code is internally coupled with neutron kinetic code Dyn3D. • Explicit coupling of overlapped computational domains was used. • A coupled system of Athlet/Dyn3D+Fluent codes was successfully tested on a real case.
Science and Technology of Nuclear Installations, 2009
The NURESIM Project of the 6th European Framework Program initiated the development of a new-gene... more 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 applic...
<jats:p>The paper describes methods used for evaluation of pressurized thermal shock (PTS) ... more <jats:p>The paper describes methods used for evaluation of pressurized thermal shock (PTS) at UJV Rez and selected results. The UJV Rez participated on the one hand in a number of international programs focused on development of PTS methodology, benchmark testing etc. On the other hand UJV has worked in several projects evaluating PTS for nuclear power plants in the Czech Republic and abroad. The introductory part of the paper also briefly mentions a development of PTS methodology in the world, its current status and trends.</jats:p> <jats:p>The core of the paper describes thermal hydraulic part of the PTS evaluation process, which usually contains system thermal hydraulic (TH) analysis and mixing calculation. Already at the system TH calculation level, a simple 2D model of reactor downcomer (DC) is used in UJV Rez analyses — as this approach enables the prediction of 2D temperature and velocity fields in DC and improves the prediction of natural circulation flow or its stagnation in individual loops. For the subsequent mixing calculation the Computational Fluid Dynamics (CFD) computer codes are used nowadays. Comparison of 2D temperature fields from 2D system TH analysis and CFD calculation shows surprisingly good agreement for most cases. The paper shows such comparison for several representative cases and takes a more detailed look into one of them.</jats:p>
Houille Blanche-revue Internationale De L Eau, 2009
Ce rapport decrit une simulation numerique d'experiences determinees a flux calorifique criti... more Ce rapport decrit une simulation numerique d'experiences determinees a flux calorifique critique effectuees sur une boucle experimentale "Large Water Loop ". Les calculs ont ete effectues a l'aide du programme NEPTUNE_CFD. "Large Water Loop" (LWL) est une boucle a pression hydraulique inactive dont les parametres technologiques et calorifiques sont appropries pour des reacteurs nucleaires a eau sous pression. Un dispositif experimental visant a rechercher un flux calorifique critique dans un faisceau de tiges electriquement chauffees fait partie du LWL. Les conditions d'une ebullition critique sont fixees a pression constante, a temperature constante de l'eau a l'entree et a debit constant, le rendement calorifique du faisceau augmentant peu a peu. Les tiges sont des tubes creux profiles a echauffement direct des parois. Les resultats des calculs de quatre experiences typiques sont presentees dans le rapport. On a pu noter a chaque fois un acc...
This paper presents an attempt to use multiphase CFD code for prediction of the Departure from Nu... more This paper presents an attempt to use multiphase CFD code for prediction of the Departure from Nucleate Boiling (DNB) type of Critical Heat Flux (CHF). Numerical simulations of DNB in boiling flow in vertical tube were performed with the NEPTUNE_CFD V2 code. This code can simulate multicomponent multiphase flow by solving three balance equations for each phase or fluid component. A new set of validated models of physical phenomena in boiling bubbly flow was used in the calculations. Simulated cases were based on data from the Standard tables of CHF in pipes, produced by the Russian Academy of Sciences. It was found out that local DNB criterion based on void fraction equal to 0.8 does not work well with the new set of physical models implemented in NEPTUNE_CFD V2 code. But it was discovered that the criterion for DNB prediction can be based on the ratio of evaporation heat flux and total wall heat flux. Evaporation heat flux is calculated by the extended Kurul and Podowski wall boili...
This paper presents numerical simulations of the boiling flow in a tube with a Departure from Nuc... more This paper presents numerical simulations of the boiling flow in a tube with a Departure from Nucleate Boiling type of critical heat flux (CHF). Standard tables of CHF produced by the Russian Academy of Sciences were used as a data set. The simulations were performed with the multiphase code NEPTUNE_CFD V1.0.7. A simple criterion based on the void fraction at the wall was used for the CHF prediction. Four data series were selected from the tables. In every series, one of the following parameters was variable: the local equilibrium quality, the mass flux, pressure and the tube diameter. The remaining three parameters were fixed. In every data point, a numerical simulation was performed so as to find out the interval of the wall heat fluxes at which the boiling crisis occurs. NEPTUNE was able to quite accurately predict CHF in cases with high mass fluxes and high pressures. On the other hand, in one low-mass-flux case, the CHF in the calculation occurred at a wall heat flux as low as ...
This paper presents a CFD simulation of selected "Large Water Loop" critical heat flux ... more This paper presents a CFD simulation of selected "Large Water Loop" critical heat flux experiments. Calculations were performed by NEPTUNE_CFD code. The Large Water Loop (LWL) is non-active pressurised-water equipment with technological and thermal parameters corresponding to those of PWR. The CHF experimental facility (a part of the Large Water Loop) has been designed for research into CHF in water flow through a bundle of electrically heated vertical rods. The critical conditions were determined under constant pressure, inlet water temperature and mass flux and for quasi steady-state -by gradually increasing the heat input. The rods are modelled by hollow tubes with direct heating of the wall. Thirteen CHF tests were calculated with NEPTUNE. In all calculated tests, a sudden rise of the wall temperature was observed. Simulations of cases with a higher mass flux were successful. Simulation of cases with a low mass flux indicates that the modelling approach might not be su...
This paper presents simulations of the convective boiling flow performed with NEPTUNE_CFD and FLU... more This paper presents simulations of the convective boiling flow performed with NEPTUNE_CFD and FLUENT codes. The DEBORA experiments carried out at CEA Grenoble were used as an experimental data set. In these experiments, freon R12 flows upwards inside a vertical pipe. Radial profiles of the flow variables are measured at the end of the heated section. Seven DEBORA cases were selected for simulation. NEPTUNE_CFD code was used without modifications because it contains all necessary models. In FLUENT, an important part of the models has been implemented by programming in User Defined Functions. The comparison of the radial profiles of void fraction, liquid temperature, gas velocity and mean bubble diameter at the end of the heated section shows that both codes can provide reasonable results in boiling conditions. The presented work was carried out within the 6 th Framework EC NURESIM project. NEPTUNE_CFD code is implemented in the NURESIM platform.
This paper presents coupled system of thermal hydraulic code Athlet, neutron kinetic code Dyn3D a... more This paper presents coupled system of thermal hydraulic code Athlet, neutron kinetic code Dyn3D and CFD code Fluent. The coupled system is intended for simulation of complex transients such as main-steam-line-break scenarios, which cannot be modeled separately first by the system and neutron kinetic code and then by CFD code, because of the feedbacks between the codes. The presented method is limited to scenarios with single-phase flow in CFD domain. In the first part of this paper, the coupling method is described. Explicit coupling of overlapped computational domains is used in this work. Exchange of the variables at the coupling interfaces is presented as well as the implementation of the coupled system on the Linux system. The master program that controls both Fluent and Athlet/Dyn3D codes was developed. The second part of the paper presents the demonstration simulation performed by the coupled system of Fluent and Athlet/Dyn3D. “Opening of Steam Dump to the Atmosphere” test car...
Multiphase Science and Technology
Predictions of local boiling flow processes leading to departure from nucleate boiling (DNB) cond... more Predictions of local boiling flow processes leading to departure from nucleate boiling (DNB) conditions are considered. The work was performed within the Nuclear Reactor Simulations project (NURESIM) within the Sixth European Framework program. This paper focuses on the Reynolds-averaged Navier-Stokes (RANS) approach as being the most reliable for simulation of realistic bubbly flows. New physical models developed within the NURESIM project are presented and tested on various single-channel boiling experiments, differing in geometry, working fluid, and operating conditions. The applicability of the model for boiling in fuel rod bundles under industrial conditions has been demonstrated.
La Houille Blanche, 2009
ABSTRACT This report presents a CFD simulation of selected "Large Water Loop" c... more ABSTRACT This report presents a CFD simulation of selected "Large Water Loop" critical heat flux experiments. Calculations were performed by NEPTUNE_CFD code. The Large Water Loop (LWL) is non-active pressurised-water equipment with technological and thermal parameters corresponding to those of pressurized water reactors. The CHF experimental facility (a part of the Large Water Loop) has been designed for research into CHF in water flow through a bundle of electrically heated vertical rods. The critical conditions were determined under constant pressure, inlet water temperature and mass flux and for quasi steady-state - by gradually increasing the heat input. The rods are modelled by hollow tubes with direct heating of the wall. Calculation results of four typical tests are presented in this paper. In all calculated tests, a sudden rise of the wall temperature was observed. Simulations of cases with a higher mass flux were successful. Simulation of cases with a low mass flux indicates that the modelling approach might not be suitable for lower mass fluxes. The results show that NEPTUNE has potential for predicting boiling flow up to CHF in the geometry of nuclear reactor fuel assembly. The presented work was carried out as part of the NURESIM project. NEPTUNE_CFD code is implemented in the NURESIM platform. Ce rapport décrit une simulation numérique d'expériences déterminées à flux calorifique ritique effectuées sur une boucle expérimentale "Large Water Loop". Les calculs ont été effectués à l'aide du programme NEPTUNE_CFD. "Large Water Loop" (LWL) est une boucle à pression hydraulique inactive dont les paramètres technologiques et calorifiques sont appropriés pour des réacteurs nucléaires à eau sous pression. Un dispositif expérimental visant à rechercher un flux calorifique critique dans un faisceau de tiges électriquement chauffées fait partie du LWL. Les conditions d'une ébullition critique sont fixées à pression constante, à temperature constante de l'eau à l'entrée et à débit constant, le rendement calorifique du faisceau augmentant peu à peu. Les tiges sont des tubes creux profilés à échauffement direct des parois. Les résultats des calculs de quatre expériences typiques sont présentées dans le rapport. On a pu noter à chaque fois un accroissement rapide de la température des parois. Les simulations des expériences à fort écoulement ont été couronnées de succès. Les calculs des expériences à faible écoulement ont montré que le procédé indiqué ne convenait pas pour des faibles écoulements. Les résultats montrent que le programme NEPTUNE présente un réel potentiel pour les calculs d'écoulement avec ébullition ou point critique d'ébullition dans la géométrie d'un caisson de chauffage d'un réacteur nucléaire. Cette étude a été réalisée dans le cadre du projet européen NURESIM. Le programme NEPTUNE_CFD fait partie de la plateforme NURESIM.
Annals of Nuclear Energy, 2014
ABSTRACT The prediction of condensation in the presence of noncondensable gases has received cont... more ABSTRACT The prediction of condensation in the presence of noncondensable gases has received continuing attention in the frame of the Severe Accident Research Network of Excellence, both in the first (2004-2008) and in the second (2009-2013) EC integrated projects. Among the different reasons for considering so relevant this basic phenomenon, coped with by classical treatments dated in the first decades of the last century, there is the interest for developing updated CFD models for reactor containment analysis, requiring validating at a different level the available modelling techniques. In the frame of SARNET, benchmarking activities were undertaken taking advantage of the work performed at different institutions in setting up and developing models for steam condensation in conditions of interest for nuclear reactor containment.
Nuclear Engineering and Design, 2015
ABSTRACT This paper presents a benchmark performed in the frame of the SARNET-2 EU project, deali... more ABSTRACT This paper presents a benchmark performed in the frame of the SARNET-2 EU project, dealing with momentum transfer between a real-scale PWR spray and the surrounding gas. It presents a description of the IRSN tests on the CALIST facility, the participating codes (8 contributions), code-experiment and code-to-code comparisons. It is found that droplet velocities are almost well calculated one meter below the spray nozzle, even if the spread of the spray is not recovered and the values of the entrained gas velocity vary up to 100% from one code to another. Concerning sensitivity analysis, several ‘simplifications’ have been made by the contributors, especially based on the boundary conditions applied at the location where droplets are injected. It is shown here that such simplifications influence droplet and entrained gas characteristics. The next step will be to translate these conclusions in terms of variables representative of interesting parameters for nuclear safety.
Recently, the safety analyses of VVER and PWR reactors have dealt with the possibility of reactiv... more Recently, the safety analyses of VVER and PWR reactors have dealt with the possibility of reactivity-induced accidents related to the penetration of a water slug with low boron concentration into the reactor core. Loop seals at the reactor coolant pump (RCP) suction are the most likely places for the formation of these slugs. The slug is formed in the loop when there is neither natural nor forced circulation. When the circulation is restored, the slug travels towards the reactor and causes an insertion of positive reactivity in the core. This report deals with a CFD simulation of the most dangerous event - the start-up of the first RCP. Only several seconds are needed for slug to reach the core and the operator has no time for corrective action. Mixing of slug on its way to the core can reduce the danger of core recriticality. The primary objective of this study was to find out whether the FLUENT 6 CFD code is capable of predicting the mixing in the cold leg, downcomer and lower ple...
Nuclear Engineering and Design, 2014
ABSTRACT This article presents a custom condensation model for commercial CFD code Fluent. The co... more ABSTRACT This article presents a custom condensation model for commercial CFD code Fluent. The condensation model was developed for the species transport model in Fluent code and it is suitable for both compressible and incompressible flow of air–steam mixture with additional non-condensable gases. The condensation model consists of two parts: condensation in volume and condensation on the wall. Condensation in volume is modeled by “return to saturation in constant time scale” method. Condensation on the wall is calculated from diffusion of steam through a layer of non-condensable gases near the wall. The performance of the condensation model was tested on the CONAN experiments. In these experiments, air–steam mixture flows downwards through a vertical channel with square cross section. One vertical wall of the channel is cooled and the steam condenses on it. The same model was then applied in simulation of PANDA Test 9bis experiment with condensation. In this test, two vessels connected with a pipe were filled with air; and steam was released into the first vessel. As the steam concentration increased in the vessels, the steam started condensing on the walls. The results of CFD simulations of both CONAN and PANDA experiments compared well with the measured data.
Nuclear Engineering and Design, 2007
The EU project FLOMIX-R was aimed at describing the mixing phenomena relevant for both safety ana... more The EU project FLOMIX-R was aimed at describing the mixing phenomena relevant for both safety analysis, particularly in steam line break and boron dilution scenarios, and mixing phenomena of interest for economical operation and the structural integrity.This report will focus on the computational fluid dynamics (CFD) code validation. Best practice guidelines (BPG) were applied in all CFD work when choosing
The NURESIM Project of the 6th European Framework Program initiated the development of a new-gene... more 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 applic...
Nuclear Engineering and Design, 2014
h i g h l i g h t s • Coupling interface between CFD code Fluent and system code Athlet was creat... more h i g h l i g h t s • Coupling interface between CFD code Fluent and system code Athlet was created. • Athlet code is internally coupled with neutron kinetic code Dyn3D. • Explicit coupling of overlapped computational domains was used. • A coupled system of Athlet/Dyn3D+Fluent codes was successfully tested on a real case.