Turbulent Coolant Flow Research Articles

A numerical model to investigate the irradiation effect on the dynamics of nuclear fuel bundles

The severe operation conditions in nuclear reactors, such as highly turbulent coolant flow, high temperatures and excessive irradiation doses, can compromise the structural integrity of nuclear fuel bundles. Specifically, the irradiation doses influence the microstructure of the material in terms of segregation and precipitation in addition to swelling. This leads to changes in the mechanical properties of the material and consequently, the dynamic characteristics will be altered. The current study presents a fully-coupled, three-dimensional, numerical, structural model that takes into account the irradiation effects on fuel bundle material. A constitutive model that describes the thermal and irradiation effects on the mechanical properties is utilized. This constitutive model is integrated in an in-house finite element code that is capable of predicting the dynamics of fuel bundles including the nonlinear material, contact dynamics, and various fluid and acoustic excitation forces. This model is utilized to predict the response of a CANDU fuel bundle under turbulence excitation, motion-dependent forces, and acoustic pressure pulsations generated by the heat transport pumps. The fuel-to-fuel and fuel-to-pressure tube contact dynamics are also included using the single point contact method (SPC). The dynamic response of the fuel bundle along with the associated impact forces and work rates are predicted over a range of flow velocities that are typical to those in an actual CANDU power plant. The results indicate that the location of the fuel channel inside the calandria tube may exacerbate the mechanical stresses on the fuel bundle endplates, which could compromise their structural integrity.

Development and Testing of a Methodology for Assessing of the Correlation Velocity Measurements’ Accuracy for the Hydrodynamic Investigations of the Turbulent Coolant Flow in Nuclear Reactor Elements

The correlation method of the coolant flow measuring is widely used in research practice including for studying of turbulent coolant flows in scale models of elements of nuclear power plants. The aim of this work was to develop a technique for assessing the effect of noise recorded by a measuring system on the flow rate readings obtained using the correlation method.A technique to assess the effect of noise as well as the relative position and acquisition period of sensors is presented. An insignificant concentration of a salt solution (NaCl or Na2SO4 ) is used as a passive impurity which creates a conductivity gradient of the medium recorded by a conductometric system. Turbulent pulsations at the interface between two concurrent isokinetic flows in a channel with a square cross section are used as the signal source for the correlational algorithm.Paper presents the values of the turbulence′s transport time between spatial conductometers, the results of estimating the spectral power density and band of the recorded signal and also the signalto-noise ratios of the measuring system obtained on their basis which are subsequently used to estimate the confidence interval of the transport time.As a result of measurements the relationship between the confidence interval value and the signal length were obtained. The measurements which were carried out at different relative positions of conductometers make it possible to make a conclusion about an increase in the spectral width of the signal and, as a consequence, a decrease in the length of the confidence interval with increasing of distance between sensors.The presented work is an approbation of this approach for its application as part of an experimental model of a nuclear reactor in order to determine per-channel flow rates in the channels of the core simulator using mesh conductometric sensors taking into account the effect of noise.

Application of the Correlation Velocity Measurements for Hydrodynamic Investigations of Turbulent Coolant Flow in Nuclear Reactor Elements

The method of correlation measurement of the coolant flow rate, widely used for operational diagnostics of nuclear power plants, can be extensively used in research practice. The aim of this work was to apply a correlation method based on the conductometric measurement system with wire-mesh sensors for measuring a coolant flow rate.Insignificant concentration of a salt solution (NaCl or Na2SO4 ) creates a gradient of the conductivity in the flow, which is used as a passive scalar measured by the system. Authors used turbulent pulsations at the interface of two concurrent flows with identical velocities in a square channel as a signal source for the correlation method. The paper presents the methodology of the tests, test facility description, signalto-noise ratio estimation, the results of digital signal processing and comparison of the measured velocities in the model with the flowrate‒averaged velocity determined by the use of flowmeters. The measured velocity values give acceptable agreement for the turbulent flow modes. It was shown that the measurement accuracy drops sharply for low-Reynolds flows.The obtained results were used for flowrate measurements in core-imitator channels of the nuclear reactor test model. The presented paper is an approbation of this approach for its application as part of an test model of a nuclear reactor in order to determine the each duct flow rates in the channels of the core simulator using wire mesh sensors.

Further Details of a Numerical Analysis on the Thermal Hydraulic Effect of Wrapped Wire Spacers in Fuel Bundle

Abstract The application of relatively simple and cheap wrapped wire spacer in the European supercritical water-cooled reactor (SCWR) (high-performance light water reactor (HPLWR)) has been proposed in order to provide enhanced heat transfer in the fuel assembly without unacceptable penalty in pressure loss. The wires cause twisting flow in the fuel assembly, which means the coolant not only flows straight in the axial direction but also has a significant transverse velocity component, and strong mixing between neighboring subchannels occurs. The aim of this ongoing research is to numerically investigate the effect of wrapped wire spacers on thermal hydraulics of the turbulent coolant flow and its heat transfer in a small bundle of four fuel rods. One bare and six-wired geometries with varying wire pitches (1–6 turn(s) of wires) have been studied. It was found that the wires generate significant amount of transverse velocity, decrease the wall temperature, and increase the heat transfer coefficient mostly in corner subchannels which were the hottest in bare geometry. Thus, the presence of wires enhances heat transfer where it is most needed. Temperature hot spots with moderate values have been identified on the cladding wall of fuel rods. Based on the results, a technically optimal choice of number of wire turns from thermal hydraulic sense has been proposed.

Longitudinal Displacement of a Fuel Rod Nonfixed in the Support Grid under the Conditions of Hydromechanical Vibration of VVER-440 Reactor Fuel Assemblies

Experiments aimed at studying the effect of fuel rod vibration in a turbulent coolant flow has on the effort of moving and longitudinal displacement of a fuel rod in the system of VVER-440 reactor fuel assembly (FA) spacer grids (SGs) for the case when the fuel rod is not fixed in the lower support are carried out. It has been found that the fuel rod longitudinal displacement force becomes essentially (by approximately 30%) higher in a coolant flow than it is in stagnant water. The fuel rod’s longitudinal displacement force decreases with an increase in the water flow velocity and, hence, with an increase in the fuel rod bundle vibration intensity. At the water flow velocity equal to approximately 3.8 m/s, which is close to the its nominal value in the VVER-440 reactor FAs, the displacement force decreases by approximately 10% as compared with its value at the velocity equal to 0.8 m/s. The regularities relating to the change with time in the fuel rod’s longitudinal displacement speed and acceleration testify that the force resisting its motion in the spacer grid system has a periodic pattern. During vibration (flexural oscillations) of a fuel rod bundle, the stiffness in the fuel rod–SG cell system changes its value, and the axes of SG cells shift relative to one another for the characteristic period of one vibration cycle. It is particularly the above-mentioned effects connected with fuel rod bundle vibration that lead to a growth in the fuel rod’s longitudinal displacement force. The results from the accomplished calculated assessments show that the algebraic sum of the Archimedean force, gravity force, force due to pressure loss across the fuel bundle, and inertial forces with the water velocity values up to 4.81 m/s do not exceed 6% of the minimal fuel rod longitudinal displacement force. In turn, this means that the system of honeycomb-type spacer grids reliably secures the VVER-440 reactor fuel rods that are not fixed in the FA lower support grid.

Numerical analysis on the thermal hydraulic effect of wrapped wire spacer in a four rod fuel bundle

Abstract The application of relatively simple and cheap wrapped wire spacer in the European supercritical-water-cooled reactor has been proposed in order to provide enhanced heat transfer in the fuel assembly without unacceptable penalty in pressure loss. The wires cause twisting flow in fuel assembly, which means the coolant flows not only straight in the axial direction but has a significant transverse velocity component and strong mixing between neighbouring sub-channels occurs. The aim of this ongoing research is to numerically investigate the effect of wrapped wire spacers on thermal hydraulics of turbulent coolant flow and its heat transfer in a small bundle with four fuel rods. One bare and six wired geometry with different wire pitches (1–6 turn(s) of wires) have been studied. Unstructured volume mesh of tetrahedral elements with fine near wall boundary mesh of prism elements for fluid domain and structured mesh of hexahedral elements for solid domains of fuel rods, wires, etc. have been used to discretize the computational domain. After an extensive mesh and turbulence model sensitivity study, the M4 mesh resolution and the most advanced BSL Reynolds Stress turbulence model were selected for the further CFD calculations. The validation of the model with 3 turns of wires has been performed. The wires generate significant amount of transverse velocity and decrease the wall temperature most in corner sub-channels which was hottest in bare geometry. The heat transfer coefficient has been increased also most in corner sub-channels by 60–145% depending on the number of turns of wires. The other two types of sub-channel showed moderated (wall type) or negligible (central type) decrease in wall temperature and enhancement in heat transfer. Temperature hot spots with moderate values have been identified on the wall of fuel rods. The optimal technical choice from thermal hydraulic sense seems to be three turns of wires due to the pressure drop is only 18% higher than in bare geometry but maximum wall temperature is 32 °C lower while the average heat transfer coefficient is 95% higher where it is most needed, in the corner sub-channels. For these thermal benefits, the 18% penalty in the pressure drop seems to be acceptable. The outcome of this technical optimization needs further confirmation by an economic analysis while the possible long term effect of revealed hot spots on integrity of fuel rod cladding has to be investigated as well.

Comparative Estimation of the Effect from Using Different Coolants in Panel-Type Radiators of Spacecrafts

A radiant cooling heat exchanger is one of key components in a closed-cycle power installation and is the bulkiest structural part of a spacecraft. The most well-elaborated version of a cooler, known as a panel-type radiator (PR), is made according to the process arrangement of radiating panels. Selecting the optimal coolant is one of important issues in designing a PR. A liquid metal coolant in the form of molten Na–K mixture is presently regarded to be the most preferred one for these purposes. It features thermal stability, resistance to radiation, and a very high thermal conductivity. The main negative feature of Na–K melt is its explosion hazard when exposed to air, a circumstance due to which difficulties are encountered in experimentally perfecting the PR design under on-land conditions. We consider high-temperature organic coolants as an alternative to a liquid metal coolant. The aim of this study was to compare the effectiveness of using different coolants for the class of PR systems whose properties and geometrical characteristics are close to the prototypes developed at the State Scientific Center (SSC) Keldysh Research Center from composite materials on the basis of carbon fibers with high thermal conductivity. To this end, a mathematical model and relevant calculation procedure have been developed. The results from the performed calculations testify that, in view of a low specific heat of liquid metal coolant, its mass flow-rate should be a factor of 2–2.5 higher than the flowrate of high-temperature organic coolant, which entails essential loss of energy for pumping. Thus, the use of high-temperature organic coolants is more preferable for a certain class of PRs with parameters close to the considered ones. Turbulent flow of coolant is an important condition, due to which significant requirements are posed to its viscous characteristics. A diphenyl mixture can be regarded as the most efficient high-temperature coolant for the considered class of PRs.

Heat Exchange in Turbulent Flow. Part 1. Turbulent Prandtl Number

Knowledge of the characteristics of heat-exchange intensity in a turbulent coolant flow is one of the main problems of safety validation of nuclear power facilities. The maximum temperature of fuel elements and other structures is determined on the basis of the heat exchange coefficients. However, most computational relations of empirical or semi-empirical character have in many cases an error ±(15–20)%, which is unsatisfactory given the current requirements of the design of nuclear power facilities. The present review of the investigations of the turbulent boundary layer for different media performed in our country and abroad in 1950–2016 makes it possible to identify a different approach to the analysis of heat exchange intensity. The first part of the review examines the structural characteristics of the boundary layer and the evaluation of the turbulent Prandtl number $$ {\mathrm{P}}_{{\mathrm{r}}_{\mathrm{T}}} $$ for different media.

MODELING OF AN ADVANCED HEAT EXCHANGE UNIT WITH MICROCHANNELS FOR A COMBINED PHOTOENERGY SYSTEM

Purpose. Mathematical modeling of the heat exchange unit main parameters for photoenergy system based on general models with forced circulation of heat transfer fluid. Methodology. To determine the coefficient of heat transfer at a given coolant temperature and surfaces temperature necessary to determine the temperature gradient in the wall of the heat exchanger. Temperature gradients can be determined by solving the equation of energy, which depends on the distribution of the flow rate in the flow. In general, a solution of convective heat transfer fluid to flow along the plane comes to solution of the system of differential equations. Results. In the paper features of the selection of theoretical basis and mathematical modeling of thermal processes in the heat exchange unit for combination photoenergy system are presented. As a result of the simulation conducted we improve and develop high-efficiency heat exchange unit with microchannels. Testing of the proposed unit proved its high efficiency through the implementation of turbulent flow of coolant with heat transfer coefficient at 18 kW/(m 2 ×K). Analytical testing of the heat exchanger allowed showing that heat exchanger unit provides a stable operating temperature at less than 50 °C with the coolant flow rate is less than 0.3 m/s. Originality. Novelty of the proposed heat exchanger is in the optimal design of microchannels to improve the heat transfer coefficient. Practical value. The use of this heat exchanger will improve the quality and uniformity of cooling solar panels and reduce energy costs for circulation of fluid.

Numerical prediction of flow induced vibrations in nuclear reactor applications

Flow induced vibration (FIV) plays an important role in nuclear industry. In nuclear reactors, the FIV are caused by a strong interaction between the fuel rods and the turbulent coolant flow around these rods. Numerical prediction of these strongly coupled fluid structure interaction (FSI) has been a challenge and is the main focus of this work. In this article, three aspects of FSI problems are numerically studied. In this first part, two different coupling schemes namely IQN-ILS and Guess-Seidel are thoroughly assessed for a laminar flow single rod experiment performed by Vattenfall (Lillberg et al. 2015). As a next step, a turbulent flow single rod experiment is selected to assess the capabilities of two different RANS models, i.e. a linear k-ωSST and a non-linear Reynolds Stress Model. Lastly, an application based experiment of Chen and Wambsganns (1972) is selected to validate the combined effect of the selected RANS model along with the coupling method.

Numerical simulation of cooling performance of an exhaust gas recirculation (EGR) cooler using nano-fluids

A numerical model is developed to predict the performance of an exhaust gas recirculation (EGR) cooler using nanofluid as the coolant. The model accounts for turbulent flow of coolant and hot smokes on an integrated computational domain. Thermal and hydrodynamic behavior of four nanofluids comprising water as the base fluid and SiO2, TiO2, Al2O3 and Cu nanoparticles, were compared over a wide range of Reynolds numbers and various particle concentrations. The accuracy of predictions was verified by experimental data available in the literature. The Al2O3-water nanofluid was found to provide the greatest heat transfer enhancement. Quantitatively, Al2O3-water nanofluid with a volume fraction of 5% and Reynolds number of 5000 improves the heat transfer coefficient by about 16% compared to pure water. However, it was found that the heat transfer enhancement was achieved at the expense of increased pressure drop due to greater viscosity of nanofluids compared to the base fluid. It was also found that the effectiveness of nanofluids in improving the heat transfer rate decreases as the Reynolds number increase.

Закономерности формирования гидродинамических нагрузок на пучки ТВЭЛов в турбулентном потоке теплоносителя в ТВС ВВЭР

Закономерности формирования гидродинамических нагрузок на пучки ТВЭЛов в турбулентном потоке теплоносителя в ТВС ВВЭР

Experimental studies of the dynamical characteristics of fuel-element bundles of VVER fuel assemblies in turbulent coolant flow

Experimental studies of the dynamical characteristics of fuel-element bundles of VVER fuel assemblies in turbulent coolant flow

Vibrations of fuel-element bundles in VVER fuel assemblies excited by turbulent coolant flow

Complex experimental studies of hydrodynamically excited vibrations of fuel-element bundles have been performed on full-scale models of VVER-440 fuel assemblies. Pressure pulsations were used as the main hydrodynamic characteristic of turbulent coolant flow according to the effect of its load on the fuel-element surfaces in the flow. Realizations of the random hydrodynamic loads causing vibrations of fuel elements are obtained on the basis of multi-point measurements of pressure pulsations along the perimeter of the outer row of the fuels elements in a bundle.

Studying the vibration and random hydrodynamic loads on the fuel rods bundles in the fuel assemblies of the reactor installations used at nuclear power stations equipped with VVER reactors

Random hydrodynamic loads causing vibration of fuel rod bundles in a turbulent flow of coolant are obtained from the results of pressure pulsation measurements carried out over the perimeter of the external row of fuel rods in the bundle of a full-scale mockup of a fuel assembly used in a second-generation VVER-440 reactor. It is shown that the turbulent flow structure is a factor determining the parameters of random hydrodynamic loads and the vibration of fuel rod bundles excited by these loads. The results from a calculation of random hydrodynamic loads are used for estimating the vibration levels of fuel rod bundles used in prospective designs of fuel assemblies for VVER reactors.

Detection of knock occurrence in a gas SI engine from a heat transfer analysis

This paper focuses on the transient thermal signal around an engine cylinder in order to propose a new and non-intrusive method of knock detection. Numerical simulations of unsteady heat transfer through the cylinder and inside the coolant flow are performed. The amplification of wall heat transfer due to knock is taken into account by a self-developed program. It enables one to fix the instantaneous heat flux on the internal side of the cylinder. The transient component of the thermal signal is then calculated and analysed. The effect of a machining, consisting of a transverse groove, which reduces the cylinder liner thermal resistance, is investigated. This groove disturbs the turbulent coolant flow, and its shape mainly influences the heat transfer from the wall to the fluid. The numerical results show that the use of a square groove would multiply by three the amplitude of the transient thermal signal recorded within the water compared to a smooth wall case. Moreover, the amplitude can be enhanced by using a nanofluid as coolant because of its higher thermal diffusivity.

Experimental investigations of correlations between neutron power and fuel element outlet temperatures of SUPER PHENIX 1

The commissioning procedure of the 3000 MWth SUPER PHENIX 1 reactor provided for the first time the opportunity to extensively study neutron power and outlet temperature noise phenomena at a large fast breeder reactor. The noise measurements cover power levels between 15 and 100 % nominal power. Two ex-vessel and one in-vessel neutron chamber were used for the neutron noise measurements. The temperature noise was measured using standard CrAl thermocouples as well as intrinsic SS-Na thermocouples at a few SA outlets. The measurements have shown that at SPX1 there is very little correlation between neutron power and outlet temperature noise. The reasons for this have been identified by a detailed analysis of noise data. It is concluded that the major part of the outlet temperature noise is not related to the heat generation process in the SA's. It is presumably produced by turbulent coolant flow in the upper plenum. As a consequence, heat transfer parameters of the fuel elements could not be obtained from noise measurements. They had to be determined by the analysis of the outlet temperature response to fast transients of reactor power (scram). The predominant temperature noise component generated outside of the fuel elements will make it extremely difficult to detect local loss of cooling in a SA or developing fuel failure at an early stage by monitoring outlet temperature noise characteristics. However, the results also indicate that the possibilities for early failure detection by outlet temperature noise analysis could be improved significantly by placing the TC's closer to the SA outlets. The response time of CrAl thermocouples could be derived from power spectral densities of CrAl and SS-Na thermocouple noise.

F0riction losses in coolant flow through kidney shaped gundrill shank

SUMMARY The factor of factional losses for both laminar and turbulent coolant flow through the passages in a gundrill has been investigated. A kidney shaped passage in the tip has been used because it is superior to both single circular and double circular passages. An extensive experimental investigation revealed a considerable deviation from the Blasius relation for circular passages. The concept of hydraulic diameter used to predict friction in circular passages fails to predict the friction in the kidney shaped passage. The results of this investigation suggest that by introducing a concept of a characteristic dimension of the passage cross-section, the friction can be predicted in a relatively simple manner similar to the Blasius approach. Computer fitting of the experimental data resulted in an empirical relation which can be used to calculate the factor of frictional tosses in a kidney shaped gundrill passage for coolant.

Hydraulic drag in turbulent coolant flow in a porous cable

The hydrodynamics and heat transfer in the cooling of an electrical cable by radical coolant filtration are studied.

Modelling of Coulomb damping and wear of vibrating systems

Coulomb damping is often relied upon to control the response of nuclear boiler tube banks acted upon by fluctuating eddy forces from turbulent coolant flow. The required energy dissipation is affected by relative sliding movements between the clamped joints of the tube and support structure. These movements Z are in general related to the centre span deflection y which is coupled to the exciting forces by the relation Z= ∑ n=1 ∞ α ny n where the α n are termed coupling coefficients. Specific models of the damping are developed for direct Coulomb damping ( n = 1) and second-order damping ( n = 2) under conditions where a constant friction force, i.e. static loading, is assumed to act. The effects of partial slip across the joint are examined, and this gives rise to a damping contribution at amplitudes that lie below the threshold y p for gross slip at the joint. In general direct Coulomb damping gives rise to a non-linear type of damping and also changes natural frequency with amplitude. Second-order damping tends asymptotically to a constant loss factor 1 Q r at large amplitudes but in general would be comparable with the hysteretic damping. The penalty from frictional damping is the occurrence of an associated wear rate which can be expressed by dW dt = k μ dξ D dt where K is the specific wear rate, ξ D is the energy dissipated as wear and μ the friction coefficient. Since K and μ are defined by environmental and material properties compatible with nuclear environments, the design should be optimized to minimize the energy dissipation.