Thermal-hydraulic Analysis Research Articles

Thermal hydraulic and material analysis of upgraded flat-type Graphite divertor mock-up for Pakistan Spherical Tokamak (PST)

The design of a divertor is critically dependent on managing power deposition, erosion effects, and plasma configuration. An upgraded double-null divertor configuration has been developed for the Pakistan Spherical Tokamak (PST), featuring graphite targets that are actively cooled and designed to withstand a peak heat flux of 0.3 MW/m² at a pressure of 0.1 MPa. This paper presents a comprehensive thermal-hydraulic and material analysis for the upgraded flat-type divertor mock-up system, covering aspects such as material surface heat load, peak temperature rise (∆T °C) on the mock-up, and surface temperature increase in the cooling channel. The analysis includes total deformation (mm) and equivalent strain for the inner vertical target (IVT), outer vertical target (OVT), and dome structure, along with material comparison. The recommended SST K-ω turbulence model is utilized in the pressure-based transient analysis, with an inlet velocity of 1.5 m s-1 and an inlet temperature of 16.8 °C. A comparative study of the material and thermal-hydraulic analyses was performed using CFD and RELAP5. The findings reveal that graphite is more suitable than tungsten for the PST's upgraded divertor system, demonstrating its effectiveness as the preferred surface material to address heat enhancement challenges in the PST.

Thermal-hydraulic analysis of He-Xe gas mixtures in triangular rod bundle wrapped with different shaped wire wraps

Thermal-hydraulic analysis of He-Xe gas mixtures in triangular rod bundle wrapped with different shaped wire wraps

Thermal hydraulic analysis of a high-temperature gas reactor steam generator after loss of feed water accidents

The helical coil steam generator is widely employed in the energy conversion process of high-temperature gas-cooled reactors for its high thermal efficiency in producing superheated steam within the helical tubes. These tubes necessitate materials specifically designed to endure high temperatures, prevent ruptures, and ensure durability under various operational conditions. This study aims to identify the maximum temperatures arising from potential transients in the secondary system and to propose immediate safety measures for minimizing associated risks. The RELAP5/MOD3.4 code was used to simulate the behavior of the next generation nuclear plant steam generator's conceptual design using steady-state conditions. Following the validation of the model, the established steady-state conditions were utilized to simulate accident scenarios, including the feed water line break and the loss of feed water flow to the steam generator. The transient scenarios assumptions include closed flap valves and retained hot helium in the shell side. The simulation tasks were conducted both with and without activating the depressurization system, in order to assess its effect on the temperature of tubes on the wall side. The analysis results indicate that tube wall temperatures can reach 720 °C at the steam generator outlet region and 620 °C at the weld region. The immediate activation of the depressurization system with a controlled relief valve can effectively reduce these temperatures by 239 °C in the steam generator outlet and by 133 °C in the weld region. This also lowers the steam generator's secondary side pressure. As a result, an upgrade and enhancement of operational procedures is straightforward.

Thermal-hydraulic analysis of wetted airside plain fin flat tube heat exchangers and the role of vortex generators

This study addresses the demand for efficient and compact heat exchange solutions in industrial applications. Unlike previous studies that predominantly focused on round tube configurations, this research investigates the impact of vortex generators and geometrical parameters on plain-fin flat-tube designs. Using ANSYS Fluent, the study evaluated the effects of fin geometry, vortex generators, and flow conditions on thermal–hydraulic performance. The key findings revealed that vortex generators enhanced heat transfer by 5–48% by promoting fluid mixing and turbulence but also leading to an increment in friction losses reaching 35–62%, depending upon the number of vortex generators. Condensation rates increased downstream of vortex generators due to enhanced mixing with cooler surfaces. In terms of thermal–hydraulic performance, specific plain-fin configurations outperformed others, offering efficient heat transfer with relatively lower friction losses, even when compactness was considered. Some configurations were recommended for applications where heat exchanger size is less important, but high heat transfer is crucial. Lastly, it was discovered that the effect of vortex generators is more pronounced for lower to intermediate pumping powers per unit area. New empirical correlations were also derived to predict thermal–hydraulic behavior based on geometry and vortex generator placement. A comparison with the literature revealed significantly higher performance as compared to convex/concave vortex generators. Lastly, it was discovered that the effect of vortex generators is more pronounced for lower to intermediate pumping powers per unit area. Empirical correlations for “j” and “f” were correlated with all geometric parameters.

Thermal-hydraulic analysis of once-through steam generator with annular narrow slot tube

Integrated small modular reactors (SMR) have garnered significant attention in the field of nuclear energy, with the high-performance integrated steam generator being a critical technology for their advancement. The high-performance once-through steam generator (HP-OTSG) featuring an annular narrow slot tube, which leverages narrow channel enhanced heat transfer technology, holds significant potential for future applications. However, research on the design and experimental studies related to this type of steam generator remains significantly limited. This paper establishes a thermo-hydraulic analysis method for the OTSG based on the segmented heat exchanger design ideas, and the analysis method is verified by comparing the calculation results with those in the literatures. Then, the experimental studies conducted in the annular narrow slot tube and the related conclusions are comprehensively reviewed and subjected to an accurate assessment. The design parameters of a typical SMR, 200 MW Nuclear Heating Reactor-II (NHR200-II), are selected to evaluate and analyze the accuracy and validity of the correlation based on the annular narrow slot tube in the different heat transfer regions including the onset of nucleate boiling (ONB) point, saturated boiling heat transfer coefficient (hsat), dry-out point, and the post-dryout heat transfer coefficient (hpdo). Finally, the influences of operational and geometrical parameters on the performance of the HP-OTSG are analyzed, and the design optimization of the HP-OTSG is performed according to several performance indicators including Power Density (PD), Area Density (AD), pressure drop (Δp), secondary steam temperature (T2out), heat transfer factor (j), the proportion of nucleate boiling zone length in H (PNL), the point of dry-out (Xdryout), and the length of heat exchange tube (L). The model developed in this paper extends the application range of high-performance steam generators to annular narrow slot tubes. The calculation model including both of the analysis method and the heat transfer correlations established in the present study can serve as a useful tool for the design and optimization of HP-OTSG, which further prompts the development of integrated SMR with built-in steam generator.

Steady-state thermal–hydraulic analysis of an NTP reactor core based on the porous medium approach

Nuclear thermal propulsion (NTP) is a promising advanced technology which has attracted wide attention in recent years. The reactor core is an essential component of an NTP system and the corresponding thermal–hydraulic analysis is necessary. In this study, the porous medium approach was applied to the simulation of a two-pass NTP reactor core which consists of the porous prismatic cermet fuel elements. The thermodynamic property models of hydrogen and the fuel element materials were implemented, as well as the empirical correlations of the heat transfer coefficient and the friction factor. The three-dimensional simulation of a single fuel element was carried out and the results were compared against another code. The code-to-code comparison verified the applicability of the porous medium approach. The three-dimensional model of the two-pass NTP reactor core was established and the steady-state simulation was carried out. The distribution patterns of the parameters are determined by the thermal–hydraulic characteristics of the reactor core, including the nonuniform heat release, contact heat conduction and folded-flow scheme. The full-core heat-flow adaptability analysis is realized, which provides a reference for the thermal–hydraulic safety analysis of the NTP reactor.

Thermal-hydraulic performance analysis and multi-objective optimization of a microchannel with staggered semi-elliptical ribs

Thermal-hydraulic performance analysis and multi-objective optimization of a microchannel with staggered semi-elliptical ribs

Validation of DDC-3D code system for neutronics and thermal-hydraulics coupling analysis using BEAVRS benchmark

The direct whole-core calculations can provide accurate results and insights to the physics phenomena of the reactor. It can also capture the local effects of temperature and density fields on fuel depletion. However, the computational cost of the direct whole-core calculations is expensive. To compromise between computational cost and accuracy, the DDC-3D code system has been developed to perform neutronics and thermal-hydraulics coupling analysis. The DDC-3D code system couples the open-source codes DRAGON & DONJON based on two-step method and subchannel code COBRA-EN. The Picard iteration method is applied to ensure the stability of numerical calculation. Then the BEAVRS benchmark is used to validate the computational capabilities of DDC-3D code system. The critical boron concentrations, control rod worths and fission rate distributions are calculated in HZP condition. The results show a good agreement with measured data. The results demonstrate that the two-step method is applicable and valid for multiphysics simulations. For the result of HFP condition for cycle 1, the results also agree well with measured data, including the trend of the critical boron concentrations and power distributions throughout the cycle 1. Although the detailed thermal–hydraulic experimental values are not available, the thermal-hydraulics analysis of the hot fuel assemblies indicates that the calculation results are reasonable. In general, the results demonstrate the feasibility and accuracy of DDC-3D code system for neutronics and thermal-hydraulics coupling calculations and life cycle simulation of PWRs.

Numerical investigations on the performance of sc-CO2 sequestration in heterogeneous deep saline aquifers under non-isothermal conditions

CO2 sequestration in geological formations is an efficient step towards mitigating climate change by reducing global warming. The CO2 emitted from natural sources and anthropogenic activities is injected into geological formations. The flow and transport of CO2 during sequestration operations require in-depth knowledge of the effects of the type of geological formation. The present work analyses the CO2 migration and storage in a heterogeneous deep saline aquifer with variable porosity and evolving permeability under non-isothermal flow conditions while upscaling the capillary pressure using the Leverett J-function. The developed numerical model is history-matched with CO2 gas saturation profiles reported using analytical and numerical approaches. The novelty of the present work lies in its ability to capture fine-scale heterogeneities in the aquifer, which is modelled through variation in the porosity (ϕ) distribution and the permeability (k). The thermal-hydraulic numerical analysis projected a pronounced effect of the degree of heterogeneity arising from the variation in porosity and permeability on pressure buildup along the top of the formation decreasing from a maximum of about 6.41 MPa at the injection well to about 0.933 MPa at 1200 m and affecting gas characteristics and transmission. The thermal front could only propagate to a maximum extent of about 142 m from the injection well in aquifers, while gas plume lengths migrated to a maximum plume length of about 1010 m. The permeability and porosity of the aquifer are critically observed to vary during the sequestration by a ratio (k/k0) of about 1.07 to 1.01 and a percentage of about 0.791 % – 3.136 % in the low permeable conditions. Maximum effective storage of CO2 by a factor of 0.95 observed in low-permeable heterogeneous aquifers with normally distributed porosity affirms maintenance of optimum gas injection rates below the fracture gradient in these formations to sequester the CO2 economically.

Transient thermal hydraulic analysis of a small passive lead–bismuth eutectic cooled fast reactor

The lead–bismuth eutectic cooled fast reactor (LFR) emerges as a cutting-edge technology in nuclear reactor design. This study introduces a compact, passive, and long-life LFR reactor LFR-180. The LFR-180 incorporates modular helical-coiled once-through steam generators (H-OTSG) and direct reactor auxiliary cooling systems. This integration enables the generation of superheated steam and passive safety features of the LFR-180, eliminating the necessity for a water-steam separation device and contributing to enhanced power generation efficiency. A detailed transient thermal hydraulic analysis is conducted with tailored models for LBE solidification and H-OTSG thermal hydraulics. Results demonstrate that LFR-180 is a fully passive reactor, requiring no intervention up to 149.1 h after shutdown. The H-OTSG can serve as the heat exchangers of the active safety systems that can effectively reduce the peak core outlet temperature from 1003.6 K under the SBO accident to 724.2 K after shutdown.

ARIANT Assessments for CANDU Analysis Using RD-14M Experiments

ARIANT (AlgoRIthm for Analysis of Network Thermalhydraulics) is a Canadian Nuclear Laboratories system thermal-hydraulic code for the modeling and analysis of two-phase flow and heat transfer for pressurized heavy water reactors, light water pressurized water reactors, and advanced reactor applications. This paper presents ARIANT models and simulations of RD-14M experiments, including small-break loss-of-coolant accidents, large-break loss-of-coolant accidents, loss-of-flow accidents, station blackout, and natural circulation, that are representative of accident scenarios in a CANDU reactor. ARIANT predictions of pressures, flow rates, temperatures, and void fractions are compared against the steady-state and transient data over the course of the tests. The results show that ARIANT predicted the key parameters with reasonable accuracy, as well as the overall behavior of the five transient events. These assessments support ARIANT’s applicability to the corresponding CANDU design-basis accidents and demonstrate ARIANT as an alternative to existing system thermal-hydraulic codes for CANDU safety analysis.

Experimental study on thermal-hydraulic performance of helically coiled tubes steam generator of lead-bismuth reactor

The advantages of the helically coiled tube steam generator lie not only in its high heat transfer efficiency but also in its compact structure, which is initially used in Pb–Bi–SCO2 reactors, and its main function is to transfer the heat from the lead-bismuth side to the SCO2 side. In this experiment, two sets of helical heat exchanger tubes are selected as the objects of investigation. Experiments are conducted on the heat transfer performance of a helical tube to study the influence of structural parameters, mass flux, pressure, and heat flux on the heat transfer characteristics during subcooled boiling and post-dryout. Considering the results of the study, new heat transfer characteristic correlations are derived for the subcooled boiling and post-dryout heat transfer regions, which provide data and theoretical support for the engineering application and thermal hydraulic analysis of steam generators.

Validation of a three-dimensional two-phase code, CUPID, using a rod-bundle boiling test facility, SIRIUS-3D

In nuclear power plants, the occurrence of a three-dimensional mixing phenomenon in two-phase flow can help mitigate the peak cladding temperature of fuels in a core. Simulating this phenomenon necessitates subchannel-scale analysis. CUPID is a three-dimensional thermal hydraulic analysis code developed by KAERI. This paper presents the validation results of the code against rod bundle tests conducted using SIRIUS-3D, a 5 × 5 rod bundle test facility. One test utilized subchannel void sensors, while the other employed X-ray CT. To validate the SIRIUS-3D test data, subchannel-scale nodes were constructed for CUPID calculation. For improved convergence, the implicit friction calculation was conducted with a small CFL ratio. The drag coefficient interfacial friction model without the EVVD model yielded the best prediction. The void fraction was over-predicted at a low velocity of 0.1 m/s, which could be improved by reducing the interfacial friction.

Effect of inclination on SB LOCA transient of a floating nuclear reactor in MARS-KS analysis using moving reactor model

As the imperative to reduce Green House Gas emissions grows within the maritime sector, nuclear power emerges as a promising solution. Especially, offshore floating nuclear power plants have garnered significant attention and development efforts from various institutions and nations. These types of reactors provide unique benefits such as extended refueling periods for advanced safety and exceptional mobility, facilitating access to remote regions. However, the external forces are induced by operating on a moving condition, that can impact the thermal–hydraulic characteristics of various hydraulic components in the entire reactor system. Addressing the safety issues of offshore floating reactors requires conducting thermal–hydraulic analyses under diverse oceanic motions. To achieve this objective, this study centered on analyzing accidents in an offshore floating reactor, specifically the BANDI-60, targeting the Small Break Loss of Coolant Accident induced by a double-ended guillotine break in the Direct Vessel Injection line. This study utilizes the system thermal–hydraulic code MARS-KS moving reactor model considering various tilting conditions and a BANDI-60 nodalization with a quarter-divided reactor pressure vessel and containment vessel volumes in the circumferential direction. The inclination conditions include static inclination along the x and y directions. Significantly, static inclination, particularly when the break aligned downward with the inclination direction, emerged as the most critical condition affecting reactor core integrity. Additionally, the containment vessel volume affects the equilibrium pressure after a break occurs, with smaller volume resulting in smaller peak cladding temperature values in static inclined conditions.

Three-dimensional CFD analysis of PHWR exposed core under postulated severe accident condition

A three-dimensional steady-state thermal–hydraulic analyses of a large sized Pressurized Heavy Water Reactors (PHWRs) with varying levels of exposure is carried out for degraded core heat transfer assessment and presented in this paper. Low-frequency postulated severe accident without mitigating actions leads to exposure of fuel channels in PHWRs. Such exposure of fuel channels leads to their heat-up, governed by interdependent heat transfer mechanisms such as radiation, convection and conduction heat transfer and thermo-chemical interactions among the core constituents. In order to understand such complex behavior and plan the severe accident management guidelines, numerical analyses of the exposed core are of necessary. This paper illustrates the influence of the temperature and velocity profiles vis-à-vis different regions of the exposed core on the fuel channel heat-up. The interesting interdependency of convection and radiation heat transfer on fuel channel heat-up is reported. The dominance of convection heat transfer is observed for initial exposures of the fuel channels. However, the advanced stages of fuel channel exposures are observed to be dominated by radiation heat transfer. Intriguingly, it is observed that the convective cooling of exposed fuel channels by steam flow is counter-acted by exothermic heat generation from Zircaloy-made calandria tube surface and steam interaction. This leads to supplemental power addition for fuel channel heat-up. Moreover, it is interesting to note that the heat-up of fuel channels is restricted to central regions of the exposed core due to degraded steam cooling and higher-powered fuel channels. It is worth noting that fuel channel disassembly is observed in the advanced levels of fuel channel exposures. Moreover, fuel channel length equivalent to nearly 7–8 fuel bundle lengths undergo disassembly depending upon the fuel channel elevation and location in the core. The calandria vault water is observed to undergo film boiling at the onset of fuel channel disassembly.

Design and thermal-hydraulic analysis of multi-target system with 100 MeV proton linear accelerator for the production of 67Cu and 68Ge radioisotopes

Radioisotopes are widely used in the fields of medicine, science, and industry. The growing demand for medical radioisotopes has driven research on alternative production methods. In particular, both isotopes of 67Cu and 68Ge play vital roles in the medical environment in many countries to be used in the radio-immunotherapy and the positron emission tomography imaging, respectively. This study designed a multi-target system consisting of two Zn and one Ga2O3 plates to enable simultaneous production of the medical radioisotopes 67Cu and 68Ge using 100 MeV proton beams. To understand the thermal effect on the multi-targets, we examined the distribution of energy absorbed in each solid plate target when exposed to an accelerated proton beam through the thermal-fluid analysis based on ANSYS simulation. For confirming thermal stability for two Zn targets and one Ga2O3 target, the modified water flow path inside the multi-target system was designed effectively with the controlled distribution of multiple sub-holes between main inlet and the following four channels. It was confirmed that the newly designed multi-target system of Zn and Ga2O3 solid plates shows higher thermal stability than the case of uniform distribution of water inlet, which means it could be exposed to a higher current beam of 7.57% to decrease the processing time.

Thermal-hydraulic analysis of a full-scale lead-bismuth cooled fast reactor core considering inter-wrapper flow

The Lead-bismuth cooled Fast Reactors (LFR) utilizes wrapped hexagonal fuel assemblies arranged in a honeycomb pattern. The inter-wrapper flow (IWF) between adjacent assemblies has a significant influence on the core thermal-hydraulics. In this paper, the LFR core assembly analysis model, inter-wrapper flow and heat transfer model, and multi-scale coupling analysis model were developed and were implemented into the previously self-developed subchannel-level three-dimensional thermal hydraulic analysis code CorTAF-LBE for whole core of LFR, which is based on the OpenFOAM computational fluid dynamics simulation platform. The inter-wrapper flow and heat transfer model was validated against the KALLA-IWF experiment. Whole core steady state operating condition and low flow rate condition simulations were performed with reference to the 19-rod bundle geometry from KALLA laboratory and typical LFR core assembly arrangement. The whole core steady state distribution laws of key thermal-hydraulic parameters were obtained, and the influence of low flow rate operation on safety parameters such as cladding and pellet temperatures was analyzed. Under low flow rate operating conditions, the IWF heat transfer was accounted for 4.87% of the core thermal power, which was 315.6 kW higher than that under normal operating conditions, playing an important role in core heat removal and safe operation. This work provides important insights into LFR core design and the assessment of the role of IWF under low flow rate conditions.

Thermal hydraulic and structural analysis of the IFMIF-DONES liquid-lithium target system

In the IFMIF-DONES neutron source a single accelerator line delivers 125 mA of deuterons at 40 MeV to a flowing liquid lithium target with a free surface facing the deuteron beam.Neutrons with fusion-relevant spectrum are generated by a stripping reaction and used for material samples irradiation.This report presents a numerical thermal–hydraulic and structural analysis using updated design of the target system and neutronics analysis results. Simulation scope includes the target assembly, vacuum chamber with beam ducts, outlet channel, quench tank vessel and connecting elements.Simulations have been conducted with the commercial CFD code Simcenter Star-CCM + . The computational domains consist total of 47.4x106 fluid and 14x106 structure cells.The following aspects have been analyzed:•Temperature distribution in lithium and target system structure during beam-on operation;•Transient mixing of lithium in the quench tank with temperature change during the beam switch-on phase.•Thermal displacements and stresses in the TA and QT structures.Simulations using the new distribution of nuclear heating show no significant changes in the temperature field in liquid lithium. Calculated unwanted temperature hotspots in the target structure demonstrate the need for optimization of the target design.

Research of VDT scheme for porous media thermal-hydraulics analysis of plate-type fuel assembly

Thermal-hydraulics analysis of nuclear reactor core is crucial for the safe operation of nuclear power plants. Porous media models in computational fluid dynamics (CFD) code are commonly used to simulate flow and heat transfer in nuclear reactor cores because they simplify geometry and reduce computational cost. However, their low spatial resolution limits detailed analysis of internal flow and temperature fields. This research develops a Virtual Domain Technology (VDT) scheme to improve the resolution of porous media models without increasing mesh density. VDT sets localized resistance coefficients to create virtual solid and fluid domains within the porous core representation. The flow and thermal effects of these virtual domains mimic real structural effects. VDT was verified on a cylinder flow case and showed excellent agreement with CFD for velocity and pressure fields. Application to a plate-type research reactor core demonstrated VDT’s ability to capture detailed temperature profiles and flow traces consistent with CFD, using only 6% as many mesh cells. By improving spatial resolution, VDT enhances porous media-based modeling of reactor cores to provide more accurate thermal-hydraulics analysis, while maintaining computational efficiency. This will aid optimization and safety analysis for advanced reactor designs.

Application of BPNN algorithm in thermal-hydraulic analysis of unwrapped LFR core

Due to the complex structure of the LFR fuel rod bundle, refined CFD simulation consumes significant computing resources and time, making it difficult to meet the immediate needs of engineering design and research. This paper introduces a method that integrates the sub-channel level thermal-hydraulic analysis code CorTAF-LBE with the BPNN algorithm to analyze the thermal-hydraulic phenomena within the LFR core. The method has been validated by refined simulation results of the THEADES test section, which indicate that the error is within an acceptable range. The feasibility of its application in LFR fuel assemblies was also evaluated. The flow characteristics and temperature distribution in unwrapped LFR rod bundles are investigated. Due to differences in fuel distribution and flow area, the temperature in the central channel is much higher than that in the side and trapezoidal channels. Three accident conditions, including inlet blockage, center blockage, and outlet blockage, were simulated. The corresponding peak temperatures of the fuel cladding were 670 K, 767 K, and 860 K, respectively, representing an increase of approximately 170 K compared to the nominal conditions. The analysis results effectively demonstrate the 3D temperature distribution of the fluid in the rod bundles, which is significant for reactor design and safety analysis of LFRs.