The High Performance Light Water Reactor (HPLWR) is one of the six innovative nuclear energy systems proposed by the Generation IV International Forum. The use of water at supercritical pressure as the coolant in the HPLWR allows a significative increase of the thermal efficiency of the power plant, a reduction in size and complexity of the system and a safety improvement with respect to the use of two phase flows. Fluids at supercritical pressure are characterized by a sharp change of properties, which may lead to an enhancement or deterioration of their heat transfer properties, whose underlying mechanisms are mainly driven by buoyancy and acceleration effects. The motivation of this research is therefore to understand the effect of the sharp change of properties in the fluid flow structure and turbulence production. This work focuses on the influence of buoyancy in particular. The effect of the strongly varying properties, which are far beyond the so-called Boussinesq approximation, was experimentally studied in a water-filled, cubical Rayleigh-Benard cell using Particle Image Velocimetry (PIV). A temperature difference of 40 K is imposed between the bottom and top plate of the cell, ensuring non-Boussinesq conditions. These experiments were conducted at Rayleigh and Prandtl numbers of 6.8 x 108 and 4.4, respectively. For the first time in literature, the instantaneous and averaged flow structures under non-Bousinesq conditions have been experimentally determined on a cross section of the whole domain. Results reveal a slight asymmetric behavior of the fluid due to the large temperature difference between the bottom arid the top plates of the cell, which is a sign of non-Boussinesq effects. The data provided in this study can be used to gain a more in depth understanding of the effect of the strongly varying proprieties of supercritical fluids on natural convection phenomena in supercritical water cooled reactors.@en

Few-group cross sections used in nodal calculations derive from standard energy collapsing and spatial homogenization performed during preliminary lattice transport calculations, that implicitly assume an infinite array of identical fuel-assemblies. The infinite-medium neutron flux used for cross section weighting does not account for environmental effects arising in case of heterogeneous configurations, which can lead to considerable leakages out of or into the assembly and thus invalidate the reflective boundary conditions used for the lattice simulation. Core-environment effects can also cause variations, with respect to the infinite-lattice calculation, in the reference cross section distribution used for few-group constant collapsing. These sources of inaccuracy prevent from reproducing with high fidelity the best estimate of the reaction rates and multiplication factor coming from the reference transport global solution. Rehomogenization techniques are therefore needed. The purpose of the present paper, which builds upon previous work done at AREVA in the area of rehomogenization, is to formalize a mathematical model that encompasses the different kinds of homogenization errors. In order to investigate the accuracy of the corresponding cross section corrections, numerical tests of an assembly-configuration sample are presented.@en

Heated or cooled fluids at supercritical pressure show large variations in thermophysical properties, such as the density, dynamic viscosity and molecular Prandtl number, which strongly influence turbulence characteristics. To investigate this, direct numerical simulations were performed of a turbulent flow at supercritical pressure (CO2 at 8 MPa) in an annulus with a hot inner wall and a cold outer wall. The pseudo-critical temperature lies close to the inner wall, which results in strong thermophysical property variations in that region. The turbulent shear stress and the turbulent intensities significantly decrease near the hot inner wall, but increase near the cold outer wall, which can be partially attributed to the mean dynamic viscosity and density stratification. This leads to decreased production of turbulent kinetic energy near the inner wall and vice versa near the outer wall. However, by analysing a transport equation for the coherent streak flank strength, it was found that thermophysical property fluctuations significantly affect streak evolution. Near the hot wall, thermal expansion and buoyancy tend to decrease streak coherence, while the viscosity gradient that exists across the streaks interacts with mean shear to act as either a source or a sink in the evolution equation for the coherent streak flank strength. The formation of streamwise vortices on the other hand is hindered by the torque that is the result of the kinetic energy and density gradients. Near the cold wall, the results are reversed, i.e. the coherent streak flank strength and the streamwise vortices are enhanced due to the variable density and dynamic viscosity. The results show that not only the mean stratification but also the large instantaneous thermophysical property variations that occur in heated or cooled fluids at supercritical pressure have a significant effect on turbulent structures that are responsible for the self-regeneration process in near-wall turbulence. Thus, instantaneous density and dynamic viscosity fluctuations are responsible for decreased (or increased) turbulent motions in heated (or cooled) fluids at supercritical pressure.@en

A parametric study on the stability performance of a prototypical natural circulation BWR is performed with the downscaled GENESIS facility. The GENESIS design is based on fluid-to-fluid modeling and includes an artificial void reactivity feedback (VRF) system for simulating the neutronic-thermal-hydraulic coupling. In this work a more sophisticated VRF system than its predecessors is developed and implemented. The VRF allowed investigating different configurations relevant for the reactor design. The experiments show that changing the fuel rods diameter to a half (doubling) decreases (increases) the stability performance of the system while the resonance frequency increases (decreases). In addition, it is found that the use of MOX fuels in a BWR slightly decreases the stability performance of the reactor. On top of this, it is clearly observed that at least two oscillatory modes exists in the system, the thermal-hydraulic mode (associated to density waves traveling thorough the core plus chimney section) and the so-called reactor mode (related to density waves travelling thorough the core). It is observed that the last one is amplified by increasing (in an absolute sense) the void reactivity feedback coefficient. Details regarding the interplay between these oscillatory modes is also given.

@en

From the beginning of BWR technology it was realized that a BWR can become unstable under particular circumstances caused by a feedback between the thermal-hydraulics and the neutronics. This instability can result in oscillations of the power and the flow rate, which is an unwanted phenomenon. The NACUSP project addresses the stability issues in current and future BWRs by expanding the basic understanding through well structured testing and analyses of experimental data, by analyses of existing operational stability data from three different European reactors (Forsmark, Leibstadt, Cofrentes), by applying this knowledge via efficient models and validated computer codes to operating reactors and reactor designs, and by developing general guidelines for reactor operation and design on how to avoid BWR instabilities. In order to cover the parameter range as efficiently as possible, four existing, sophisticated thermohydraulic test facilities (CLOTAIRE [Gouirand, J.M., 1988. CLOTAIRE Program, description and manufacturing of the mock-up, CEA Cadarache, DRE/STRE/LGV 88¿876.] DESIRE [van de Graaf, R., van der Hagen, T.H.J.J., Mudde, R.F., 1994. Two-phase flow scaling laws for a simulated BWR assembly. Nucl. Eng. Des. 148, 455¿462.] CIRCUS [de Kruijf, W.J.M., van der Hagen, T.H.J.J., Mudde, R.F., 2000. CIRCUS; a natural circulation two-phase flow facility, Eurotherm Seminar No. 63, 6¿8 September 1999 Genoa, Italy, 391¿395] and PANDA [Dreier, J., Huggenberger, M., Aubert, C., Bandurski, T., Fischer, O., Healzer, J., Lomperski, S., Strassberger, H.-J., Varadi, G., Yadigaroglu, G., 1996. The PANDA facility and first test results, Kerntechnik 61, 214¿222]) have been selected. To extrapolate from small-scale separate-effect testing conditions to full-scale integral reactor conditions one needs to rely on the performance of computer codes (MONA [Hoyer, N., 1994. MONA, a 7-Equation Transient two-phase flow model for LWR dynamics, Proceedings of the International Conference on New Trends in Nuclear System Thermohydraulics, pp. 271¿280], ATHLET [Krepper, E., Prasser, H.-M., 1999. Natural circulation experiments at the ISB-VVER integral test facility and calculations using the thermal-hydraulic code ATHLET. Nucl. Technol. 128, 75¿86], RAMONA-3(-5) [Grandi, G., Hoyer, N., Belblidia, L., 1998. Two-fluid thermal hydraulics capabilities of RAMONA-5, Proceedings of the International Conference on the Physics of Nuclear Science and Technology, October 5¿8, 1998, Long Island, NY, USA, 1621¿1625], LAPUR-V [Otaduy, P.J., March-Leuba, J., 1990. LAPUR User's Guide, NUREG/CR-542, ORNL/TM-11285], DWOS, FLICA [Toumi, I., Bergeron, A., Gallo, D., Royer, E., Caruge, D., 2000. FLICA-4: a three-dimensional two-phase flow computer code with advanced numerical methods for nuclear applications, Nucl. Eng. Des. 200, 139¿155], SAPHYR, RELAP5/MOD3 [Idaho National Engineering Laboratory, 1995. RELAP5/MOD3 Code Manual, Vols. 1¿6, INEL-95/0174, NUREG/CR-5535], TRAC-BF1 [Idaho National Engineering Laboratory, 1992. TRAC-BF1/MOD1; an advanced best-estimate computer program for BWR accident analysis, Vols. 1¿3, EGG-2626, NUREG/CR-4356]. For specific items CFD codes are applied as well. Within NACUSP a linear stability analysis tool is developed. Four of the three experimental facilities within the project have yielded a large, unique database. Natural-circulation and stability characteristics at nominal pressure were collected from extensive experiments at the DESIRE facility. Low-pressure characteristics were measured at the PANDA (large scale) and the CIRCUS (small scale) facility. The phenomena encountered (flow rate and stability trends, the occurrence of flashing induced oscillations) can be explained from basic physical models and are now well understood. Several thermal-hydraulic codes have been benchmarked¿some successfully, others with limited success¿against these data. The CLOTAIRE (nominal pressure, large scale) facility has been modified and is now ready for experiments. Nuclear power plant data from three BWRs (Cofrentes, Forsmark and Leibstadt) have been collected. A very complete set of reactor data was obtained from measurements at the Leibstadt BWR during cycle-19. During this test, which was performed within the NACUSP project, safe reactor operation was demonstrated at extreme conditions, covering the entire power-flow map. State-of-the-art, coupled thermal-hydraulic¿neutronics codes are being benchmarked on these data. The development of an easy to use, rapid and efficient analytical tool for parametric studies on BWR stability is in progress. The last year of NACUSP will be devoted to finalising the before mentioned activities and to use the tools developed and the experience obtained for developing general guidelines for designing and operating BWRs.@en

From the beginning of BWR technology it was realized that a BWR can become unstable under particular circumstances caused by a feedback between the thermal-hydraulics and the neutronics. This instability can result in oscillations of the power and the flow rate, which is an unwanted phenomenon. The NACUSP project addresses the stability issues in current and future BWRs by expanding the basic understanding through well structured testing and analyses of experimental data, by analyses of existing operational stability data from three different European reactors (Forsmark, Leibstadt, Cofrentes), by applying this knowledge via efficient models and validated computer codes to operating reactors and reactor designs, and by developing general guidelines for reactor operation and design on how to avoid BWR instabilities. In order to cover the parameter range as efficiently as possible, four existing, sophisticated thermohydraulic test facilities (CLOTAIRE [Gouirand, J.M., 1988. CLOTAIRE Program, description and manufacturing of the mock-up, CEA Cadarache, DRE/STRE/LGV 88¿876.] DESIRE [van de Graaf, R., van der Hagen, T.H.J.J., Mudde, R.F., 1994. Two-phase flow scaling laws for a simulated BWR assembly. Nucl. Eng. Des. 148, 455¿462.] CIRCUS [de Kruijf, W.J.M., van der Hagen, T.H.J.J., Mudde, R.F., 2000. CIRCUS; a natural circulation two-phase flow facility, Eurotherm Seminar No. 63, 6¿8 September 1999 Genoa, Italy, 391¿395] and PANDA [Dreier, J., Huggenberger, M., Aubert, C., Bandurski, T., Fischer, O., Healzer, J., Lomperski, S., Strassberger, H.-J., Varadi, G., Yadigaroglu, G., 1996. The PANDA facility and first test results, Kerntechnik 61, 214¿222]) have been selected. To extrapolate from small-scale separate-effect testing conditions to full-scale integral reactor conditions one needs to rely on the performance of computer codes (MONA [Hoyer, N., 1994. MONA, a 7-Equation Transient two-phase flow model for LWR dynamics, Proceedings of the International Conference on New Trends in Nuclear System Thermohydraulics, pp. 271¿280], ATHLET [Krepper, E., Prasser, H.-M., 1999. Natural circulation experiments at the ISB-VVER integral test facility and calculations using the thermal-hydraulic code ATHLET. Nucl. Technol. 128, 75¿86], RAMONA-3(-5) [Grandi, G., Hoyer, N., Belblidia, L., 1998. Two-fluid thermal hydraulics capabilities of RAMONA-5, Proceedings of the International Conference on the Physics of Nuclear Science and Technology, October 5¿8, 1998, Long Island, NY, USA, 1621¿1625], LAPUR-V [Otaduy, P.J., March-Leuba, J., 1990. LAPUR User's Guide, NUREG/CR-542, ORNL/TM-11285], DWOS, FLICA [Toumi, I., Bergeron, A., Gallo, D., Royer, E., Caruge, D., 2000. FLICA-4: a three-dimensional two-phase flow computer code with advanced numerical methods for nuclear applications, Nucl. Eng. Des. 200, 139¿155], SAPHYR, RELAP5/MOD3 [Idaho National Engineering Laboratory, 1995. RELAP5/MOD3 Code Manual, Vols. 1¿6, INEL-95/0174, NUREG/CR-5535], TRAC-BF1 [Idaho National Engineering Laboratory, 1992. TRAC-BF1/MOD1; an advanced best-estimate computer program for BWR accident analysis, Vols. 1¿3, EGG-2626, NUREG/CR-4356]. For specific items CFD codes are applied as well. Within NACUSP a linear stability analysis tool is developed. Four of the three experimental facilities within the project have yielded a large, unique database. Natural-circulation and stability characteristics at nominal pressure were collected from extensive experiments at the DESIRE facility. Low-pressure characteristics were measured at the PANDA (large scale) and the CIRCUS (small scale) facility. The phenomena encountered (flow rate and stability trends, the occurrence of flashing induced oscillations) can be explained from basic physical models and are now well understood. Several thermal-hydraulic codes have been benchmarked¿some successfully, others with limited success¿against these data. The CLOTAIRE (nominal pressure, large scale) facility has been modified and is now ready for experiments. Nuclear power plant data from three BWRs (Cofrentes, Forsmark and Leibstadt) have been collected. A very complete set of reactor data was obtained from measurements at the Leibstadt BWR during cycle-19. During this test, which was performed within the NACUSP project, safe reactor operation was demonstrated at extreme conditions, covering the entire power-flow map. State-of-the-art, coupled thermal-hydraulic¿neutronics codes are being benchmarked on these data. The development of an easy to use, rapid and efficient analytical tool for parametric studies on BWR stability is in progress. The last year of NACUSP will be devoted to finalising the before mentioned activities and to use the tools developed and the experience obtained for developing general guidelines for designing and operating BWRs.@en

From the beginning of BWR technology it was realized that a BWR can become unstable under particular circumstances caused by a feedback between the thermal-hydraulics and the neutronics. This instability can result in oscillations of the power and the flow rate, which is an unwanted phenomenon. The NACUSP project addresses the stability issues in current and future BWRs by expanding the basic understanding through well structured testing and analyses of experimental data, by analyses of existing operational stability data from three different European reactors (Forsmark, Leibstadt, Cofrentes), by applying this knowledge via efficient models and validated computer codes to operating reactors and reactor designs, and by developing general guidelines for reactor operation and design on how to avoid BWR instabilities. In order to cover the parameter range as efficiently as possible, four existing, sophisticated thermohydraulic test facilities (CLOTAIRE [Gouirand, J.M., 1988. CLOTAIRE Program, description and manufacturing of the mock-up, CEA Cadarache, DRE/STRE/LGV 88¿876.] DESIRE [van de Graaf, R., van der Hagen, T.H.J.J., Mudde, R.F., 1994. Two-phase flow scaling laws for a simulated BWR assembly. Nucl. Eng. Des. 148, 455¿462.] CIRCUS [de Kruijf, W.J.M., van der Hagen, T.H.J.J., Mudde, R.F., 2000. CIRCUS; a natural circulation two-phase flow facility, Eurotherm Seminar No. 63, 6¿8 September 1999 Genoa, Italy, 391¿395] and PANDA [Dreier, J., Huggenberger, M., Aubert, C., Bandurski, T., Fischer, O., Healzer, J., Lomperski, S., Strassberger, H.-J., Varadi, G., Yadigaroglu, G., 1996. The PANDA facility and first test results, Kerntechnik 61, 214¿222]) have been selected. To extrapolate from small-scale separate-effect testing conditions to full-scale integral reactor conditions one needs to rely on the performance of computer codes (MONA [Hoyer, N., 1994. MONA, a 7-Equation Transient two-phase flow model for LWR dynamics, Proceedings of the International Conference on New Trends in Nuclear System Thermohydraulics, pp. 271¿280], ATHLET [Krepper, E., Prasser, H.-M., 1999. Natural circulation experiments at the ISB-VVER integral test facility and calculations using the thermal-hydraulic code ATHLET. Nucl. Technol. 128, 75¿86], RAMONA-3(-5) [Grandi, G., Hoyer, N., Belblidia, L., 1998. Two-fluid thermal hydraulics capabilities of RAMONA-5, Proceedings of the International Conference on the Physics of Nuclear Science and Technology, October 5¿8, 1998, Long Island, NY, USA, 1621¿1625], LAPUR-V [Otaduy, P.J., March-Leuba, J., 1990. LAPUR User's Guide, NUREG/CR-542, ORNL/TM-11285], DWOS, FLICA [Toumi, I., Bergeron, A., Gallo, D., Royer, E., Caruge, D., 2000. FLICA-4: a three-dimensional two-phase flow computer code with advanced numerical methods for nuclear applications, Nucl. Eng. Des. 200, 139¿155], SAPHYR, RELAP5/MOD3 [Idaho National Engineering Laboratory, 1995. RELAP5/MOD3 Code Manual, Vols. 1¿6, INEL-95/0174, NUREG/CR-5535], TRAC-BF1 [Idaho National Engineering Laboratory, 1992. TRAC-BF1/MOD1; an advanced best-estimate computer program for BWR accident analysis, Vols. 1¿3, EGG-2626, NUREG/CR-4356]. For specific items CFD codes are applied as well. Within NACUSP a linear stability analysis tool is developed. Four of the three experimental facilities within the project have yielded a large, unique database. Natural-circulation and stability characteristics at nominal pressure were collected from extensive experiments at the DESIRE facility. Low-pressure characteristics were measured at the PANDA (large scale) and the CIRCUS (small scale) facility. The phenomena encountered (flow rate and stability trends, the occurrence of flashing induced oscillations) can be explained from basic physical models and are now well understood. Several thermal-hydraulic codes have been benchmarked¿some successfully, others with limited success¿against these data. The CLOTAIRE (nominal pressure, large scale) facility has been modified and is now ready for experiments. Nuclear power plant data from three BWRs (Cofrentes, Forsmark and Leibstadt) have been collected. A very complete set of reactor data was obtained from measurements at the Leibstadt BWR during cycle-19. During this test, which was performed within the NACUSP project, safe reactor operation was demonstrated at extreme conditions, covering the entire power-flow map. State-of-the-art, coupled thermal-hydraulic¿neutronics codes are being benchmarked on these data. The development of an easy to use, rapid and efficient analytical tool for parametric studies on BWR stability is in progress. The last year of NACUSP will be devoted to finalising the before mentioned activities and to use the tools developed and the experience obtained for developing general guidelines for designing and operating BWRs.@en

From the beginning of BWR technology it was realized that a BWR can become unstable under particular circumstances caused by a feedback between the thermal-hydraulics and the neutronics. This instability can result in oscillations of the power and the flow rate, which is an unwanted phenomenon. The NACUSP project addresses the stability issues in current and future BWRs by expanding the basic understanding through well structured testing and analyses of experimental data, by analyses of existing operational stability data from three different European reactors (Forsmark, Leibstadt, Cofrentes), by applying this knowledge via efficient models and validated computer codes to operating reactors and reactor designs, and by developing general guidelines for reactor operation and design on how to avoid BWR instabilities. In order to cover the parameter range as efficiently as possible, four existing, sophisticated thermohydraulic test facilities (CLOTAIRE [Gouirand, J.M., 1988. CLOTAIRE Program, description and manufacturing of the mock-up, CEA Cadarache, DRE/STRE/LGV 88¿876.] DESIRE [van de Graaf, R., van der Hagen, T.H.J.J., Mudde, R.F., 1994. Two-phase flow scaling laws for a simulated BWR assembly. Nucl. Eng. Des. 148, 455¿462.] CIRCUS [de Kruijf, W.J.M., van der Hagen, T.H.J.J., Mudde, R.F., 2000. CIRCUS; a natural circulation two-phase flow facility, Eurotherm Seminar No. 63, 6¿8 September 1999 Genoa, Italy, 391¿395] and PANDA [Dreier, J., Huggenberger, M., Aubert, C., Bandurski, T., Fischer, O., Healzer, J., Lomperski, S., Strassberger, H.-J., Varadi, G., Yadigaroglu, G., 1996. The PANDA facility and first test results, Kerntechnik 61, 214¿222]) have been selected. To extrapolate from small-scale separate-effect testing conditions to full-scale integral reactor conditions one needs to rely on the performance of computer codes (MONA [Hoyer, N., 1994. MONA, a 7-Equation Transient two-phase flow model for LWR dynamics, Proceedings of the International Conference on New Trends in Nuclear System Thermohydraulics, pp. 271¿280], ATHLET [Krepper, E., Prasser, H.-M., 1999. Natural circulation experiments at the ISB-VVER integral test facility and calculations using the thermal-hydraulic code ATHLET. Nucl. Technol. 128, 75¿86], RAMONA-3(-5) [Grandi, G., Hoyer, N., Belblidia, L., 1998. Two-fluid thermal hydraulics capabilities of RAMONA-5, Proceedings of the International Conference on the Physics of Nuclear Science and Technology, October 5¿8, 1998, Long Island, NY, USA, 1621¿1625], LAPUR-V [Otaduy, P.J., March-Leuba, J., 1990. LAPUR User's Guide, NUREG/CR-542, ORNL/TM-11285], DWOS, FLICA [Toumi, I., Bergeron, A., Gallo, D., Royer, E., Caruge, D., 2000. FLICA-4: a three-dimensional two-phase flow computer code with advanced numerical methods for nuclear applications, Nucl. Eng. Des. 200, 139¿155], SAPHYR, RELAP5/MOD3 [Idaho National Engineering Laboratory, 1995. RELAP5/MOD3 Code Manual, Vols. 1¿6, INEL-95/0174, NUREG/CR-5535], TRAC-BF1 [Idaho National Engineering Laboratory, 1992. TRAC-BF1/MOD1; an advanced best-estimate computer program for BWR accident analysis, Vols. 1¿3, EGG-2626, NUREG/CR-4356]. For specific items CFD codes are applied as well. Within NACUSP a linear stability analysis tool is developed. Four of the three experimental facilities within the project have yielded a large, unique database. Natural-circulation and stability characteristics at nominal pressure were collected from extensive experiments at the DESIRE facility. Low-pressure characteristics were measured at the PANDA (large scale) and the CIRCUS (small scale) facility. The phenomena encountered (flow rate and stability trends, the occurrence of flashing induced oscillations) can be explained from basic physical models and are now well understood. Several thermal-hydraulic codes have been benchmarked¿some successfully, others with limited success¿against these data. The CLOTAIRE (nominal pressure, large scale) facility has been modified and is now ready for experiments. Nuclear power plant data from three BWRs (Cofrentes, Forsmark and Leibstadt) have been collected. A very complete set of reactor data was obtained from measurements at the Leibstadt BWR during cycle-19. During this test, which was performed within the NACUSP project, safe reactor operation was demonstrated at extreme conditions, covering the entire power-flow map. State-of-the-art, coupled thermal-hydraulic¿neutronics codes are being benchmarked on these data. The development of an easy to use, rapid and efficient analytical tool for parametric studies on BWR stability is in progress. The last year of NACUSP will be devoted to finalising the before mentioned activities and to use the tools developed and the experience obtained for developing general guidelines for designing and operating BWRs.@en

Abstract: The feasibility of particle image velocimetry (PIV) in a thermally convective supercritical fluid was investigated. Hereto a Rayleigh–Bénard convection flow was studied at pressure and temperature above their critical values. The working fluid chosen was trifluoromethane because of its experimentally accessible critical point. The experiments were characterized by strong differences in the fluid density from the bottom to the top of the cell, where the maximum relative density difference was between 17 and 42%. These strong density changes required a careful selection of tracer particles and introduced optical distortions associated with strong refractive index changes. A preliminary background oriented schlieren (BOS) study confirmed that the tracer particles remained visible despite significant local blurring. BOS also allowed estimating the velocity error associated with optical distortions in the PIV measurements. Then, the instantaneous velocity and time-averaged velocity distributions were measured in the mid plane of the cubical cell. Main difficulties were due to blurring and optical distortions in the boundary layer and thermal plumes regions. An a posteriori estimation of the PIV measurement uncertainty was done with the statistical correlation method proposed by Wieneke (Measure Sci Technol 26:074002, 2015). It allowed to conclude that the velocity values were reliably measured in about 75% of the domain. Graphic abstract: [Figure not available: see fulltext.].@en

Heat transfer to fluids at supercritical pressure is different from heat transfer at lower pressures due to strong variations of the thermophysical properties with the temperature. We present and analyze results of direct numerical simulations of heat transfer to turbulent CO2 at 8 MPa in an annulus. Periodic streamwise conditions are imposed so that mean streamwise acceleration due to variations in the density does not occur. The inner wall of the annulus is kept at a temperature of 323 K, while the outer wall is kept at a temperature of 303 K. The pseudocritical temperature Tpc=307.7 K, which is the temperature where the thermophysical properties vary the most, can be found close to the inner wall. This work is a continuation of an earlier study, in which turbulence attenuation due to the variable thermophysical properties of a fluid at supercritical pressure was studied. In the current work, the direct effects of variations in the specific heat capacity, thermal diffusivity, density, and the molecular Prandtl number on heat transfer are investigated using different techniques. Variations in the specific heat capacity cause significant differences between the mean nondimensionalized temperature and enthalpy profiles. Compared to the enthalpy fluctuations, temperature fluctuations are enhanced in regions with low specific heat capacity and diminished in regions with a large specific heat capacity. The thermal diffusivity causes local changes to the mean enthalpy gradient, which in turn affects molecular conduction of thermal energy. The turbulent heat flux is directly affected by the density, but it is also affected by the mean molecular Prandtl number and attenuated or enhanced turbulent motions. In general, enthalpy fluctuations are enhanced in regions with a large mean molecular Prandtl number, which enhances the turbulent heat flux. While analyzing the Nusselt numbers under different conditions it is found that heat transfer deterioration or enhancement can occur without streamwise acceleration or mixed convection conditions. Finally, through a combination of a relation between the Nusselt number and the radial heat fluxes, a quadrant analysis of the turbulent heat flux, and conditional averaging of the heat flux quadrants, it is shown that heat transfer from a heated surface depends on the density and the molecular Prandtl number of both hot fluid moving away from a heated surface as well as the thermophysical properties of relatively cold fluid moving towards it.@en

In many industrial processes that involve heat transfer and turbulent flows, thermo-physical properties can vary significantly due to pressure and (or) temperature variations.@en