Thermal Hydraulic Investigations Research Articles

Thermal-hydraulic investigation of liquid metal cross flow different arrangements of tube bundles under pulsating

The topic of this paper is to couple the passive and active approaches to enhance the heat transfer performance in the inline tube bundle heat exchanger. Firstly, the k-kl-ω turbulence model and Kays turbulent Prandtl number model are validated by liquid metal cross flow tube bundle experiments and rectangular groove pulsating flow experimental results. Then the numerical studies were carried out for pulsating velocity amplitude of 0.5, frequency of 10, and Re of 20,000 with transverse and streamwise pitch-to-diameter ratios of 1.5, 1.65, and 1.8. The introduction of pulsating flow effectively improves the turbulence characteristics between the tube bundles, especially at the first three rows, where the velocity fluctuation is enhanced by about 1.5–2 times. The tube heat transfer performances are all improved. The performance evaluation criterion (PEC) variation ranges from 1.04 to 1.32, which signifies that the introduction of pulsating flow can enhance the comprehensive heat transfer performance. The per unit flow rate thermal entropy generation decreased by about 9.7–33.3 %, in addition, a phase difference between the transient thermal entropy production and pulsating velocity was found. The correlation equations about Nu and f factor within the scope of this paper are presented.

Thermo-hydraulic analysis of direct steam generation in a 500 m long row of parabolic trough solar collector using Eulerian two-fluid modeling approach

Direct steam generation (DSG) in the parabolic trough solar collectors is a potential alternative for economic and performance improvement of the solar thermal system. This process comprises the preheating, evaporation, and superheating of water/steam in the solar collectors. The modeling and simulation of the evaporation section are challenging because of the complex physics associated with the boiling phase change process. This work describes the Eulerian two-fluid modeling of the once-through mode of the DSG process in solar collectors. The interfacial interactions between phases have been modeled using the appropriate empirical correlations. The experimental data from the Spanish DISS test facility is used to validate the numerical model. Further, the 3-D thermal–hydraulic investigation of DSG in a 500 m long solar collectors' row has been performed for operating pressures of 60 bar and 100 bar; inlet mass flow rates of 0.4 kg/s and 0.6 kg/s; DNI of 750 W/m2 for the solar time at 12:00 h and 13:00 h. It has been observed that the maximum absorber temperature is reduced by 27 % at 60 bar and 20 % at 100 bar operating pressure when the inlet mass flow rate increases from 0.4 kg/s to 0.6 kg/s. Similarly, the maximum circumferential temperature difference reduces by 12 % and 14 %, respectively. Moreover, the fluid temperature, steam quality, absorber temperature distributions, fluid velocity, and pressure loss under the subjected boundary conditions have been summarized.

Experimental and numerical study of an innovative twined tube HX design

An innovative twined tube heat exchanger (HX) design is proposed, which is simple, robust, and efficient for use in advanced nuclear reactors. The heat exchanger is designed with two tubes twined around each other. Multiple modules can be assembled within a shell to meet the necessary heat load. Thermal-hydraulic investigations of the novel design are performed through experiments and numerical simulation. The experimental investigations of the twined tube HX are carried out on a laboratory-scale thermal–hydraulic rig commissioned at the Pakistan Institute of Engineering & Applied Sciences (PIEAS). The primary loop contains water at a high temperature and a pressure of 4.0 bar pressure, while the secondary loop operates at atmospheric pressure. Single-phase forced convection heat transfer characteristics of the twined tube HX are investigated experimentally. Pressure drop is measured for mass fluxes ranging from 350 kg/m2s to 1376 kg/m2s. A correlation for the estimation of friction factor is developed based on the experimental studies. Heat transfer measurements are carried out for mass fluxes ranging from 244 kg/m2s to 419 kg/m2s. Based on these experiments, the Nusselt number is evaluated for the mass flux range mentioned above. The numerical simulations of a twined tube HX are performed on a full-scale STAR-CCM + model. The model is validated against the experimental results. The performance factor of the twined tube HX is higher than that of straight tube HX, with a maximum performance enhancement of 60%. The thermal–hydraulic performance of the twined tube HX is comparable with helical coil HX designs. However, the proposed HX design is easier to fabricate and maintain than the helical coiled HX. Overall, the study provides valuable insights into the thermal–hydraulic performance of the twined tube HX, which could be useful in the design of compact and efficient heat exchangers for small modular reactors.

Coupled thermal-hydraulic investigation on the heat extraction performance considering a fractal-like tree fracture network in a multilateral well enhanced geothermal system

Due to their high conductivity, fractures play a vital role in geothermal exploitation. Previously, multiple random fractures, considered as the primary channel in enhanced geothermal systems (EGS), were studied extensively. However, natural fractures may not be developed in hot dry granite and fluid flow may be controlled by faults or hydraulic fractures. Based on fractal theory, this paper considers that hydraulic fractures have fractal-like tree characteristics and focuses on the influence of their geometric and heterogeneous characteristics on heat extraction. The results show that a higher fracture number, fracture length and a staggered fracture layout are the key factors affecting heat extraction, which obviously enhance the heat extraction efficiency. Additionally, a higher fracture length ratio and aperture ratio and a greater distance between the primary fracture and the lateral well are suggested in EGS, which are important factors influencing heat extraction. However, the effect is weaker than the above 3 factors. The fracture stage, the spacing between the primary fractures and the fracture angle are secondary factors, which should not be excessively focused on in EGS. The results can provide some reasonable suggestions for reservoir fracturing optimization.

Thermal-Hydraulic Stability Investigations of Once Through Helically Coiled Steam Generators for Smr Applications

Thermal-Hydraulic Stability Investigations of Once Through Helically Coiled Steam Generators for Smr Applications

Thermal-hydraulic investigation on micro heat sinks with ribbed pin-fin arrays and single heating input: parametrical study

Thermal-hydraulic investigation on micro heat sinks with ribbed pin-fin arrays and single heating input: parametrical study

Application of the best-estimate model calibration and prediction through experimental data assimilation methodology to the tests performed on a helium cooled First Wall Mock-up

Abstract Experiments are under progress at the Karlsruhe Institute of Technology (KIT) as part of the general strategy for the development of the DEMO Helium Cooled Pebble Bed (HCPB) design. These experiments are part of the program on the qualification of components and on the validation of system codes. In 2018, thermal-hydraulic investigations have been performed looking into the transient response of a helium-cooled blanket First Wall Mock-Up under Loss Of Flow Accident (LOFA) conditions. The results obtained through these investigations captured the quick temperature rise on the surfaces exposed to heat loads (such as those coming from the plasma), and provided a robust set of data for the validation of system codes such as RELAP5-3D. Then, the present work describes the validation activities performed on the RELAP5-3D code. This validation was performed employing also a Best-Estimate methodology called “Best-Estimate Model Calibration and Prediction through Experimental Data Assimilation methodology”. In a previous work, the same methodology was used to calibrate a RELAP5-3D model of the HELOKA-HP electrical heater under normal operation. In this regard, the present work marks the first application of this methodology to fast transients in the field of fusion reactors.

A CFD study on the effect of size of fuel sphere on PBR core

In this work, a thermal-hydraulic investigation of N2 as a coolant in a pebble bed reactor core has been performed using a porous media approach. Three different diameters of fuel sphere have been employed for the numerical simulations. The pebble bed reactor is a kind of packed bed reactor whose core is a long right circular cylinder with a height of 3.5 m and an outer diameter of 3.7 m. The finite volume method was used to solve the governing equations using ANSYS FLUENT 14.5. Several important thermal-hydraulic parameters have been investigated consisting of the coolant and solid temperatures, density, pressure drop, and the coolant temperature.

Thermal hydraulic investigation of heat transfer from a completely blocked fuel subassembly of SFR

A non-classical porous body model has been adopted to predict the thermal hydraulics within a totally blocked prototype spent fuel subassembly of a sodium cooled fast reactor. The heat transfer within the subassembly is by natural convection of sodium and the ultimate heat sink is inter-wrapper flow. The decay heat is considered as a source term in the energy equation. Governing equations for 3-D, steady state, porous-body based natural convection have been solved by using the commercial CFD code ANSYS WORKBENCH 15.0. The present study considers a 217 pin wire wrap bundle of spent fuel subassembly, for an axial length of five helical pitches. The magnitudes of maximum sodium temperature and natural convection velocity inside the blocked subassembly have been obtained from the model. Through parametric studies, the permissible decay power (i) that avoids boiling of sodium trapped inside the subassembly (450 kW) and (ii) that restricts the peak clad temperature less than 823 K (100 kW), have been obtained. Further, an equivalent conduction model for the natural convection has been developed which can be used for quick estimate of sodium temperature at various values of decay power. It is seen that the conductivity value of sodium in the equivalent conduction model varies as keq(q′′′) = kNa + C×(q′′′)n where the values of C and n are found to be 1 and 0.24. From the point of view of safety, subassembly outlet sodium temperature monitoring has to be initiated at this power level itself, during power raising operation in a fast reactor. The permissible decay power from clad safety point of view is reached 6 h after reactor shut-down. This suggests that fuel handling operation can be initiated ideally after this time.

Calculation of hydraulic channels of drives with taking in to account temperature and viscosity changes

The thermal hydraulic investigation of hydraulic channels is presented. Changes of pressure and flow values due to non-steady thermodynamic processes were considered. Main aim of investigation is to ensure hydraulic calculation, that had been made to predict working characteristics of hydraulic drives at design stage, for the variable temperature environment. Structural analyses of the flow circumstances along channels were made with taking in to account temperature gradient and viscosity changes. Numerical simulation had been made to find out period of velocity stabilization and its value. The investigation considers geometry and shape of channels via taking in to account pressure reduction and velocity changes in hydraulic system. Results of investigation allow to increase the precision of hydraulic calculation for drive, that works at variable temperature and none-stable modes, using proposed method.

Thermal hydraulic investigations of ALLEGRO ceramic fuel assemblies

Gas-cooled Fast Reactor (GFR) is one of the generation IV reactor concepts. Fast neutron spectrum and high outlet temperature of the coolant make this reactor type a contributor to sustainability of nuclear technology. One of the main challenges of the concept is the development of innovative fuel assemblies, which can withstand the high temperature. Because of the high temperature, the thermal conditions in the core have to be known in detail to design and operate the reactor safely.Main issue of this paper is to investigate the thermal hydraulics of inner and corner regions of the ceramic assemblies of the GFR demonstrator, ALLEGRO. First, models have been built for bare rod bundle regions to determine the inlet boundary conditions of the complex models. Based on these simulations models have been developed, which include active length ceramic rods with four spacer grids in the regions. These models describe heat convection in the active part of the rod bundles and heat conduction in the clads and assembly shroud. Using these models temperature fields in typical regions are determined and effects of the heat conduction in the clads and thermal radiation are also investigated. Considering the computational effort and accuracy, modelling of the rod cladding and shroud material is required, while modelling of the radiative heat transport is less important.

Thermal hydraulics analysis of a helical coil steam generator of a small modular reactor

In the present study, thermal hydraulic investigation of a steam generator of an integral small modular reactor (SMR) has been performed. For this investigation, a helical coil steam generator of IRIS design is selected due to availability of its design parameters in literature. Steam generator model is developed in RELAP5/Mod 4.0. At first, steady state simulation has been performed for this model and results are found in good agreement with the IRIS design data. In order to evaluate the behavior of steam generator under accidental conditions, transient analysis has also been performed. Steam generator performance is assessed for two transients, namely, feed line break (FLB) and loss of flow accident (LOFA). Exponential losses of secondary side pressures (inlet and outlet) are considered for FLB, whereas an exponential loss of primary flow rate is considered for LOFA. The results indicate safe and reliable operation of IRIS steam generator for both transients.

SIMULATE-3 K coupled code applications

Abstract This paper describes the coupled code system TRACE/SIMULATE-3 K/VIPRE and the application of this code system to the OECD PWR Main Steam Line Break. A short description is given for the application of the coupled system to analyze DNBR and the flexibility the system creates for the user. This includes the possibility to compare and evaluate the result with the TRACE/SIMULATE-3K (S3K) coupled code, the S3K standalone code (core calculation) as well as performing single-channel calculations with S3K and VIPRE. This is the typical separate-effect-analyses required for advanced calculations in order to develop methodologies to be used for safety analyses in general. The models and methods of the code systems are presented. The outline represents the analysis approach starting with the coupled code system, reactor and core model calculation (TRACE/S3K). This is followed by a more detailed core evaluation (S3K standalone) and finally a very detailed thermal-hydraulic investigation of the hot pin condition (VIPRE).

Heat transfer to liquid metals with empirical models for turbulent forced convection in various geometries

This paper gives a summary of liquid metal properties and empirical models for thermal-hydraulic investigations. Liquid metals are used as coolant in nuclear and fusion energy and concentrated solar power systems, which requires a correct representation of their thermal-hydraulic behavior. Current generations of system codes, describing the thermal-hydraulic behavior of plants or experimental facilities, include only a few metals (sodium and lead-bismuth) flowing only in circular tubes and rod bundles. Empirical heat transfer models for rectangular ducts or annular channels are usually not used in system codes used. Furthermore, many experiments for liquid metal heat transfer have been performed with metals other than sodium and lead-bismuth. Therefore, thermo-physical properties of several liquid metals are presented based on a literature review. Empirical models for various geometries and boundary conditions are presented in this paper, too. Empirical models are implemented into a system code and validated by means of post-test analyses of experiments. The verification and validation process of the properties and the empirical models shows a good agreement between experiment and prediction for all considered conditions. It also shows the difference between the investigated geometries, which justifies the provision of tailored empirical models for comprehensive investigations of liquid metal flows.

System thermalhydraulics for design basis accident analysis and simulation: Status of tools and methods and direction for future R&D

System thermalhydraulic investigations of Design Basis Accident require several tools and methods including the Process Identification and Ranking Table, the scaling, experiment analysis, modelling, code development, code Validation and Verification, and Uncertainty Quantification. This paper intends to give an overview of these methods and tools showing what the state of the art is, and presenting some recent advances. Recommendations are made with future direction for R&D including the need of new advanced experiments and instrumentation, and the future role of CFD and multi-scale analyses. For many people it is not clear what current system codes are, and what they can be. Then the main characteristics of these codes are recalled and propositions are made to clarify the code capabilities and limitations and to improve the knowledge of the conditions for a correct application of the codes for safety in a Best Estimate Plus Uncertainty approach. Also, the on-going developments of 3-field models and Transport of Interfacial Area are summarized and associated experimental needs are identified. The growing role of 3D modelling of reactor core and Pressure Vessel requires additional experimental data for a proper validation. CFD in open medium also contributes to investigations when 3D geometrical aspects play an important role. Recent activities performed in the OECD-NEA Working Group for Analysis and Management of Accidents is summarized and recent applications of two-phase CFD to boiling flows and two-phase PTS scenarios are reported. The role of a multi-scale approach for safety issues is illustrated with the LOCA transients in LWRs. Attention is focused on the need of specific validation experiments and of consolidated uncertainty methods for both system codes and CFD codes.

Thermal hydraulic investigations on porous blockage in a prototype sodium cooled fast reactor fuel pin bundle

Thermal hydraulic characteristics of sodium flow in a prototype fuel subassembly with porous internal blockage have been investigated by computational fluid dynamics (CFD) simulations. CFD solutions for a subassembly having 217 pin bundle with seven helical pitch length were obtained by parallel processing. The CFD model has been validated against benchmark blockage experiment reported in literature. Wide parametric ranges for blockage radius, porosity, mean particle diameter and location of blockage have been considered. Critical length of blockage that would result in local sodium boiling as a function of aforementioned blockage parameters has been estimated and the parametric zone posing risk of sodium boiling has been identified. Attention has been paid to coolant mixing and flow and temperature fields downstream of the blockage zone. It is seen that for a prototype subassembly with various sections contributing to pressure loss, the total flow reduction is <2.5% for all blockages that can lead to local sodium boiling. This suggests, that global bulk sodium temperature monitoring at subassembly outlet is unlikely to detect slowly growing blockages. Comparing the sodium flow and temperature fields in unblocked and blocked bundles, it is found that the wake-induced temperature non-uniformity persist even upto 3 helical pitch length, highlighting that the sodium temperature non-uniformity at the bundle exit can serve as an efficient blockage indicator, provided that the cross-section temperature is mapped by a proper instrumentation. The peak clad temperature is found to be a strong function of porosity, with enhanced clad temperature for smaller porosity. Fuel-clad that are partly exposed to blockage are subjected to large circumferential temperature variation and the resulting huge thermal stress. Further, for a six subchannel blockage with 40% porosity and 0.5mm mean particle diameter the critical length is 80mm, whereas for the same blockage the critical length reduces to <7mm when its porosity reduces to 5%. Six subchannel blockage with 60% porosity and 0.5mm mean particle diameter, does not induce boiling even up to a blockage height of 400mm. For a single subchannel blockage with one helical pitch length, there is no risk of sodium boiling even for porosity as low as 5%. The results of the present study would act as safety and monitoring criteria during the operation of the reactor.

Coarse‐Grid‐CFD for the Thermal Hydraulic Investigation of Rod‐Bundles

AbstractA nuclear reactor core, that is a few meters in height and diameter is composed of hundreds of fuel assemblies which are again composed of tenth of fuel rods with a diameter of about 10 mm. The relevant length scales for a Computational Fluid Dynamics (CFD) simulations range from the sub millimetre range, relevant for the sub channels up to several meters. Describing such a multi‐scale situation with CFD is extremely challenging and the traditional approach is to use integral methods. These are sub channel and sub assembly analyses codes requiring closure by empirical and experimental correlations. A CFD simulation of a complete nuclear reactor set up resolving all relevant scales requires exceedingly large computational resources. However, in many cases there exists repetitive geometrical assemblies and flow patterns. Based on this observation the general approach of creating a parametrized model for a single segment and composing many of these reduced models to obtain the entire reactor simulation becomes feasible. With the Coarse‐Grid‐CFD (CGCFD) ( [1], [2]), we propose to replace the experimental or empirical input with proper CFD data. Application of the methodology starts with a detailed, well‐resolved, and verified CFD simulation of a single representative segment. From this simulation we extract in tabular form so‐called volumetric forces which upon parametrization is assigned to all coarse cells. Repeating the fine simulation for multiple flow conditions parametrized data can be obtained or interpolated for all occurring conditions to the desired degree of accuracy. Note, that parametrized data is used to close an otherwise strongly under‐resolved, coarsely meshed model of a complete reactor set up. Implementation of volumetric forces are the method of choice to account for effects as long as dominant transport is still distinguishable on the coarse mesh. In cases where smaller scale effects become relevant the Anisotrop Porosity Formulation (APF) allows capturing transport phenomena occurring on the same or slightly smaller scale compared to the coarse mesh resolution. Within this work we present results of several fuel assemblies, that were investigated with our methodology. In particular, we show Coarse‐Grid‐CFD simulations including a 127 pin LBE cooled wire wrapped fuel assembly. (© 2015 Wiley‐VCH Verlag GmbH &amp; Co. KGaA, Weinheim)

Thermal hydraulic investigations and optimization on the EVC system of a PWR by CFD simulation

In order to optimize the design of Reactor Pit Ventilation (EVC) system in a Pressurized Water Reactor (PWR), it is necessary to study the characteristics of the velocity, pressure and temperature fields in the EVC system. A full computational fluid dynamics (CFD) model for the EVC system is constructed by a commercial CFD code, where the complex fluid and solid coupling is treated. The Shear Stress Transport (SST) model is adopted to perform the turbulence calculation. This paper numerically investigates the characteristics of the velocity, pressure and temperature distributions in the EVC system. In particular, the effects of inlet air parameters on the thermal hydraulic characteristics and the reactor pit structure are also discussed for the EVC system optimization. Simulations are carried out with different mesh sizes and boundary conditions for sensitivity analysis. The computational results are important references to optimize the design and verify the rationality of the EVC system.

E-SCAPE: A scale facility for liquid-metal, pool-type reactor thermal hydraulic investigations

MYRRHA (Multi-purpose hYbrid Research Reactor for High-tech Applications) is a flexible fast-spectrum research reactor under design at SCK·CEN. MYRRHA is a pool-type reactor with lead bismuth eutectic (LBE) as primary coolant.The proper understanding of the thermal hydraulic phenomena occurring in the reactor pool is an important issue in the design and licensing of the MYRRHA system and liquid-metal cooled reactors by extension. Model experiments are necessary for understanding the physics, for validating experimental tools and to qualify the design for the licensing.The E-SCAPE (European SCAled Pool Experiment) facility at SCK·CEN is a thermal hydraulic 1/6-scale model of the MYRRHA reactor, with an electrical core simulator, cooled by LBE. It provides experimental feedback to the designers on the forced and natural circulation flow patterns. Moreover, it enables to validate the computational methods for their use with LBE.The paper will elaborate on the design of the E-SCAPE facility and its main parameters. Also the experimental matrix and the pre-test analysis using computational fluid dynamics (CFD) and system thermal hydraulics codes will be described.

Thermal hydraulic investigations of an extended station blackout event in FBTR

Thermal hydraulic investigation into the capability of Fast Breeder Test Reactor (FBTR) in handling an extended station blackout event was carried out. The main concerns during this event are decay heat removal and prevention of sodium freezing in circuits. The means of removing decay heat in FBTR are (i) by natural convection of air in steam generator casing and (ii) heat loss through long sodium piping of the loop type reactor. The study is focussed towards the current operating power. The plant dynamics code DYNAM developed specific to this reactor system and validated against commissioning tests has been utilized for this purpose. It has been established that core cooling is not a concern even when the system cooling through steam generator casing is not activated. Piping heat loss is adequate to remove the decay heat. Cooling through SG casing is essential only for preventing heat up of cold leg sodium systems. Concern on sodium freezing arises when the decay power reduces to very low level below the natural heat loss through pipe lines. This situation arises only after a year for the current power level of the plant.