Thermal-hydraulic Behavior Research Articles

Small break LOCA studies for different layouts of passive safety systems in the IRIS reactor

IRIS (International Reactor Innovative and Secure) is an integral, medium power, light water reactor with advanced safety features. In the first decade of the 21st century, 22 institutions under the leadership of Westinghouse Electric Corporation were involved in its development. The University of Zagreb, along with the Polytechnic of Milan, was in charge of performing safety analyses. A detailed plant model is developed using the RELAP5 code for the analyses of thermal–hydraulic processes in the reactor vessel, the GOTHIC code for the analysis of the processes in the containment and, in addition, the ASYST code for the calculation of a severe accident. Some of the previous small break loss-of-coolant accident analyzes at the existing pipelines are repeated to test the improved plant model. However, the focus of the paper is on the new set of analyzes of hypothetical breaks along the reactor vessel with the aim of determining whether the passive safety systems can ensure successful core cooling. For this purpose, two models are developed with different configurations of the emergency heat removal system and the safety systems inside the containment that inject water into the reactor vessel. The results show the complex and rather ambiguous dependence of the reactor coolant system thermal–hydraulic behaviour on the selected boundary conditions. The scenarios analyzed vary from design basis events to severe accidents. The capabilities of specific safety systems in mitigating the consequences of an accident are determined, depending on the position and size of the break on the reactor vessel wall.

Numerical study on the hydrothermal characteristics of a wire-wrapped rod bundle with nonuniform wire pitches

In this study, thermal hydraulic behaviors in a 19-pin bundle fuel assembly with nonuniform wire pitches is investigated by combing CFD with the Kriging method. To optimize the design, two geometric variables—the ratio of inner pitch to reference pitch (Pi/P) and the ratio of outer pitch to reference pitch (Po/P)—are selected, and the design space is sampled using Latin Hypercube Sampling (LHS). The sampled points are then subjected to CFD analysis. Convergence is considered achieved when the residuals of all variables are below 1e-5. The optimization problem aims to minimize the objective function, which is a linear combination of the cross-sectional temperature difference and friction factor. Sequential Quadratic Programming (SQP) is employed to search for the optimal point using a constructed meta-model. When compared to the reference shape, the optimal shape exhibits higher axial velocity in the inner channel, higher average temperature, smaller temperature difference at the outlet section, and reduced pressure drop in the fuel assembly. The Kriging model accurately predicts the cross-sectional temperature difference and friction coefficient for the optimal shape, consistent with the CFD calculation results. This confirms the accuracy and feasibility of the Kriging model in fuel assembly optimization.

Studies of the Thermalhydraulics Subchannel Code ASSERT-PV 3.2-SC for Supercritical Applications

Abstract Over the last decade, several international thermalhydraulics benchmarking efforts have been carried out to support the development of the Generation IV supercritical water-cooled reactor (SCWR) concept. These benchmarking efforts aimed to assess the readiness of computer codes to predict the thermalhydraulics behavior of supercritical fluids for nuclear fuel assembly applications. The results from the benchmarking also shed light on knowledge gaps. Throughout the years, several advancements in this area have been achieved, resulting in relevant conclusions and observations. Furthermore, experimental campaigns have been carried out worldwide to further our knowledge on the thermalhydraulics of supercritical fluids. The nuclear industry uses the subchannel approach to study the thermalhydraulics behavior of nuclear fuel assemblies in detail. In Canada, the subchannel code advanced solution of subchannel equations in reactor thermalhydraulics—pressure velocity (ASSERT-PV) is the qualified code for subchannel applications. ASSERT-PV was modified to handle supercritical conditions, resulting in an interim code version. This publication presents relevant subchannel analyses using the interim supercritical version of ASSERT-PV for fuel assemblies cooled with supercritical fluids.

Thermal-hydraulic simulation of loss of flow accident for WWR-S research reactor

Abstract A thermal-hydraulic model has been developed by RELAP5 code to simulate the thermal-hydraulic behavior of a WWR-S research reactor under loss of flow accident (LOFA). The reactor power is 2 MW with downward flow direction and different types of fuel bundles of different power densities and different coolant flow-rates. The simulation is performed for two scenarios; protected and unprotected LOFA. In the protected LOFA scenario; Scram is triggered as soon as the coolant flow rate reaches 85 % of its nominal value while in the unprotected LOFA scenario; the reactor continues operation during pump cost-down and natural circulation flow. Once the flow is low enough, the buoyancy force increases in the core leading to flow inversion phenomenon and establishment of a natural circulation mechanism within the core coolant channels in both scenarios. In the protected LOFA scenario, the coolant remains subcooled by a vast margin during transient while bulk boiling is predicted at the upper part of the core for the unprotected LOFA scenario. The heat fluxes leading to the onset of nucleate boiling and the critical heat flux are predicted and the safety margins are determined. The results for both scenarios are analyzed and discussed.

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.

Fundamental Understanding of Marine Applications of Molten Salt Reactors: Progress, Case Studies, and Safety

Marine sources contribute approximately 2% of global energy-related CO₂ emissions, with the shipping industry accounting for 87% of this total, making it the fifth-largest emitter globally. Environmental regulations by the International Maritime Organization (IMO), such as the MARPOL (International Convention for the Prevention of Pollution from Ships) treaty, have driven the exploration of alternative green energy solutions, including nuclear-powered ships. These ships offer advantages like long operational periods without refueling and increased cargo space, with around 200 reactors already in use on naval vessels worldwide. Among advanced reactor concepts, the molten salt reactor (MSR) is particularly suited for marine applications due to its inherent safety features, compact design, high energy density, and potential to mitigate nuclear waste and proliferation concerns. However, MSR systems face significant challenges, including tritium production, corrosion issues, and complex behavior of volatile fission products. Understanding the impact of marine-induced motion on the thermal–hydraulic behavior of MSRs is crucial, as it can lead to transient design basis accident scenarios. Furthermore, the adoption of MSR technology in the shipping industry requires overcoming regulatory hurdles and achieving global consensus on safety and environmental standards. This review assesses the current progress, challenges, and technological readiness of MSRs for marine applications, highlighting future research directions. The overall technology readiness level (TRL) of MSRs is currently at 3. Achieving TRL 6 is essential for progress, with individual components needing TRLs of 4–8 for a demonstration reactor. Community Readiness Levels (CRLs) must also be addressed, focusing on public acceptance, safety, sustainability, and alignment with decarbonization goals.

Numerical investigation on the core thermal hydraulic behavior of pool-type sodium-cooled fast reactor (SFR)

A thorough understanding of the reactor core thermal hydraulic behavior is essential for the design and safety analysis of Sodium-cooled Fast Reactors (SFR). Due to the application of hexagonal subassembly, the core thermal hydraulic behavior is significantly affected by the flow field within the subassemblies, the inter-wrapper region and the hot pool. Analysis of the core thermal hydraulic behavior requires a model coupling the three regions mentioned above, which has been identified as one of the thermal hydraulic challenges in SFR. In the present study, a 3D model that covers the three regions was developed for the core of the China Experimental Fast Reactor (CEFR) with the Computational Fluid Dynamic (CFD) code, Fluent. The inter-wrapper region and the hot pool were modeled in detail, while the subassemblies were modeled with a special porous medium model. The core thermal hydraulics behavior under steady state was studied, more specifically, information for the flow field distribution at the core outlet, the inter-wrapper flow and the duct wall temperature distribution was obtained. Under steady state, liquid sodium in the inter-wrapper region is supplied by the inner region of the hot pool. And it enters the inter-wrapper region from the core outer region and returns back to the hot pool inner region from the core central region. The inter-wrapper flow is cooled by non-fuel subassemblies and heated up by fuel subassemblies. For non-fuel subassembly, the ratio of the total heat transfer rate between the inter-wrapper flow and the subassemblies to the heat generated within subassemblies could reaches 96%; for fuel subassemblies, the maximum ratio of the total heat transfer rate between the inter-wrapper flow and the subassemblies to the heat generated within subassemblies is 2.45%. Significant temperature gradients have been observed on the duct wall, with maximum values of 156.69 K/m in the vertical direction and 2,196.00 K/m in the circumferential direction. The largest temperature gradient appears on the duct of subassemblies adjacent to the transition region of fuel subassemblies and non-fuel subassemblies.

Characterizing magnetohydrodynamic effects on developed nanofluid flow in an obstructed vertical duct under constant pressure gradient

Abstract This research endeavors to conduct an examination of the thermal characteristics within the duct filled with the copper nanoparticles and water as base fluid. In exhaust systems, like car exhausts, chimneys, and kitchen hoods, duct flows are crucial. These systems safely discharge odors, smoke, and contaminants into the atmosphere after removing them from enclosed places. The study focuses on a laminar flow regime that is both hydrodynamically and thermally developed, with a specified constraints at any cross-sectional plane. To address this, we employ the finite volume method as it stands as a judicious choice, offering a balance between computational efficiency and solution accuracy. Notably, we have observed that the deceleration of flow induced by elevated Rayleigh numbers can be effectively regulated by the application of an appropriately calibrated external magnetic field. The prime parameters of the problem with ranges are: pressure gradient ( 1 ≤ p 0 ≤ 100 ) (1\le {p}_{0}\le 100) , Hartmann number ( 0 ≤ Ha ≤ 50 ) (0\le \text{Ha}\le 50) , Rayleigh number ( 1 , 000 ≤ Ra ≤ 40 , 000 ) (1,000\le \text{Ra}\le 40,000) , and magnetic parameter ( 0 ≤ M ≤ 50 ) (0\le M\le 50) . Furthermore, our analysis reveals that the Nusselt number exhibits a nearly linear correlation with the nanoparticle volume fraction parameter, a trend observed across a range of Rayleigh numbers and magnetic parameter values. We have noted that a mere 20% nanoparticle volume fraction can result in up to 62% rise in the Nusselt number while causing an almost 50% decrease in the factor f Re. This research framework serves as a robust foundation for understanding the intricate interplay between magnetic influences and thermal-hydraulic behavior within the delineated system.

Analysis of Soot Deposition Effects on Exhaust Heat Exchanger for Waste Heat Recovery System

This study investigates the thermal–hydraulic behavior and deposition characteristics of a shell and tube exhaust heat exchanger using a CFD-based predictive model of soot deposition. Firstly, considering the influences of thermophoretic, wall shear stress, and other deposition and removal mechanisms, a predictive model is developed for long-term performance of heat exchangers under soot deposition. Then, the variations in exhaust heat exchanger performance during a 4 h deposition period are simulated based on the model. Subsequently, the variation of deposition distribution and different deposition velocities are also evaluated. Finally, an analysis of the long-term performance of the exhaust heat exchanger under varying gas velocities and temperature gradients is conducted, revealing the performance variations under all engine-operating conditions. Results show that the deterioration in normalized relative j/f1/2 varies from 5.26% to 24.91% under different work conditions, and the exhaust heat exchanger with high gas velocity and low temperature gradient exhibits optimal long-term performance.

Investigation of the fast flux test facility transient behavior during a loss of flow without scram test

A Coordinated Research Project (CRP) on the benchmark analysis of Fast Flux Test Facility (FFTF) Loss of Flow Without Scram (LOFWOS) has been held under the sponsorship of the International Atomic Energy Agency. The CRP aims to improve understanding of loss of flow events in fast reactors, as well as to assess capabilities of existing computer codes against experimental data. FFTF is a 400 MWth Sodium-cooled Fast Reactor (SFR) owned by the U.S. Department of Energy, operated from 1980 to 1992. The CRP involves the LOFWOS Test #13, belonging to a series of unprotected transients performed as part of the Passive Safety Testing program. In this framework, the Nuclear Engineering Research Group of Sapienza University of Rome has developed an integrated multiphysics modelling based on a coupled approach with RELAP5-3D© (thermal-hydraulics) and PHISICS (neutron-kinetics) codes. The present paper discusses the results obtained by the authors after the participation in the benchmark open phase. As suggested by the organizers, a two-step methodology was followed. Firstly, calculations were run by imposing the reactor power as boundary condition and focusing on the FFTF thermal–hydraulic behaviour. A good accordance was detected between the numerical results and the experimental data, above all for what concerns the reference core outlet temperatures. Then, the simulations were repeated by using a coupled approach. Calculation outcomes demonstrated the capability of the selected code suite (PHISICS/RELAP5-3D®) to reproduce the reactor transient behavior and its suitability to be used as an effective numerical tool to simulate liquid metal fast reactors accidental scenarios where thermal–hydraulic and neutron-kinetic phenomena are strongly coupled.

Numerical Investigation of Air Natural Convection in the AP1000 Passive Containment Cooling System Following LBLOCA Using ANSYS FLUENT

The innovative design of the AP1000 power plant has various layers of passive safety systems aiming to enhance reactor safety during normal and transient conditions. The passive containment cooling system (PCCS) is a safety-related system capable of removing heat from the steel containment vessel (SCV) to the atmosphere and preventing the containment from exceeding the design pressure and temperature following a postulated design-basis accident. The PCCS heat removal mechanisms include condensation on the internal SCV surface, heat conduction, natural convection, evaporation of water film, and radiative heat transfer. In two basic postulated scenarios, the reactor decay heat can ultimately be removed from the SCV only by air natural convection. The first scenario occurs 72 h following a large-break loss-of-coolant accident (LBLOCA) when the passive containment cooling water storage tank becomes unavailable. The second scenario occurs following a postulated loss of shutdown decay heat removal event. Hence, investigating the thermal-hydraulic behavior of the containment under transient conditions is essential to ensure its safety and integrity. In this study, a simplified three-dimensional model using ANSYS FLUENT is developed to investigate the cooling capability of air natural convection outside the SCV during a LBLOCA event. Because of the lack of experimental data, code-to-code validation was performed using the actual results of AP1000 alongside other research findings. The results show good agreement with available data, which can be used for future research.

RPV top and bottom break SBLOCA with passive emergency core cooling system using ATLAS test facility

A SBLOCA test, named as C2.1 in the frame of the OECD/NEA ATLAS3 project, using a passive emergency core cooling system (PECCS) was performed by Korea Atomic Energy Research Institute (KAERI). In the test, two simultaneous 2-inch equivalent breaks occurred at top and bottom nozzles of the reactor pressure vessel of ATLAS test facility. The main purpose of the test is to investigate thermal hydraulic transient behavior in a reactor coolant system (RCS) during a SBLOCA scenario and to evaluate the effectiveness of PECCS. From the observation on thermal hydraulic phenomena during the test, the test period was divided into three characteristic phases such as pressure plateau, safety injection, and clad temperature re-increase phases. Continuous inventory loss of the primary loop through two break nozzles led to a primary system pressure decrease and a clad temperature increase. Even though high-pressure safety injection tanks (HPSITs) were actuated in an early stage of the pressure plateau phase, safety injection from HPSITs, however, were not actually injected. Safety injection flows from the HPSITs and SITs were practically injected to the downcomer only after the opening of automatic depressurization valves (ADVs). With a continuous decrease of a primary system pressure, low-pressure (LP) injection was started. However, LP injection could not change the increasing trend of the clad temperatures. Post-test calculations were performed using MARS-KS 1.4, and calculation results were compared with those of the test.

Performance evaluation of the SMART passive safety injection system against the SBLOCA scenario with the SMART-ITL facility

A set of system performance tests has been performed and the thermal-hydraulic behaviors were investigated for the passive safety injection system (PSIS) of the SMART design. Major thermal-hydraulic characteristics of core cooling in reactor coolant system and safety injection (SI) in PSIS were investigated using the SMART-ITL. The parametric effects of break size, break location and train number on the PSIS performance were discussed with the relevant SMART-ITL data. Compared with the 2.0 inch SI line break test, the 0.4 inch SI line break test has a milder change of system parameters. The pressure is reduced more quickly, the decrease of water level is slower and the break mass flow rate is higher during the pressurizer safety valve line break than during the SI line break. Major thermal-hydraulic parameters were compared among single, two and full train operation of the PSIS focusing on the effect of train number. It was shown that multiple trains of CMT and SIT were operated independently and the reactor vessel (RV) inventory increased proportionally with the addition of each train. Sufficient SI water was injected into the RV in a passive manner to cool down the reactor core efficiently during the full train test.

Thermal hydraulic behavior of a large dimension hybrid heat pipe/oscillating heat pipe

Similar to the layout of an oscillating heat pipe (OHP), a quartz hybrid heat pipe (HHP) with an inner diameter of 5 mm, exceeding 59.7 % of the capillary threshold of OHP, was designed. The detailed thermal hydraulic behavior of fluid in the HHP has been revealed. It was found that the HHP undergoes a heat transfer mode transition from phase change to pulsating flow. It is also means that the liquid slugs can be dynamically and stably formed if the heat input is sufficient. The single bubble growth and the multi-bubble coalescence are two ways to form the intermittent flow. In addition, as the main flow pattern in low filling ratio (FR), the formation of annular flow is closely concerned in the behaviors of fluid. Moreover, the higher the FR is, the clearer the contour of slug flow is. At the same time, the oscillating motion can be formed easily at a high FR, resulting in a better performance of HHP in lower heat load. However, as the heat input increases, HHP with a 60 % filling ratio exhibits lower thermal resistance, achieving a minimum value of 0.08 °C/W at 100 W.

A parametric optimization framework for fin-and-tube heat exchangers based on response surface methodology and artificial intelligence

This study presents a numerical investigation and provides a novel framework for the rapid parametric optimization of attack angles (α) and aspect ratios (e) in plate fin-and-tube heat exchangers (PFTHEs) featuring elliptical tubes. The methodology incorporates three-dimensional numerical simulations, response surface methodology (RSM), backpropagation neural network (BPNN), and Entropy-VIKOR. The study focuses on a PFTHE with a single elliptical tube, varying the attack angles (α) from −30° to 0° in 5° increments and the aspect ratios (e) from 0.4 to 1.0 in a step of 0.1. Air velocities (u) range from 0.4 to 3.2 m/s with an interval of 0.2 m/s, corresponding Reynolds number from 200 to 2200. The results indicate that using elliptical tubes reduces Δp and enhances heat transfer capacity, with the most significant improvements at an e of 0.4. Exponential relationships between dimensionless parameters Nu or f and Re are established for different range of α at specific e values. The synergistic effect of e, α, and Re or u on PFTHE performance is confirmed through RSM and BPNN analysis. Their R2 values are no less than 0.896 and reach a maximum of 0.998, indicating excellent predictive accuracy. Utilizing Entropy-VIKOR method, the framework identifies a suitable design solution with e = 0.6, α = -20°, and u = 3.2 m/s. This solution exhibits a 5 % increase in h and a 58 % reduction in Δp relative to the corresponding circular tube. The study provides specific expressions relating design variables to thermal–hydraulic behavior and indicates carefully selecting the α and e can improve the overall performance. In summary, these findings have significant implications for reducing optimization design period for PFTHEs with small tube diameters.

Modelling and analysis of quench in the 15-kA HTS conductor

High Temperature Superconductors (HTS) are very promising materials for possible application in future fusion magnets, with significant R&D progress on HTS conductors in recent years. However, since geometric and thermo-physical characteristics of HTS and LTS conductors differ significantly, some doubts have arisen if the approaches successfully used in numerical simulations of the thermal–hydraulic behavior of LTS conductors would be sufficient also for HTS, particularly in cases when fast transient processes (such as e.g. quench) are considered. In order to provide data for better understanding of the quench phenomenon in HTS conductors as well as for testing different numerical approaches and proper tuning of the numerical codes, a dedicated experimental campaign (Quench Experiment) was carried out at the SULTAN test facility within the international collaboration between the EUROfusion consortium and China. Our present study is a part of the work on analysis and interpretation of the data collected during this experiment. Simulations of the chosen experimental run were performed using two THEA models with different levels of complexity. Uncertain model parameters (thermal resistances and copper RRR) were explored across a wide range. Our goal was to identify the possibly simple model that accurately reproduces the experimental results.

Probabilistic analysis of seepage and temperature fields in geothermal extraction process based on discrete fracture network (DFN) model

Enhanced geothermal system is currently the most efficient technology of exploiting geothermal energy from high-temperature rock reservior in deep earth. However, due to the uncertainty of geological structural discontinuities, such as natural or artificial fractures, great challenges are encountered in conducting the precise evaluation of heat extraction within enhanced geothermal system. In this paper, a probability analysis is carried out to incorporate the uncertainty of the fracture network into the analysis of the spatial-time evolution on the seepage and temperature fields, based on a discrete fracture network based modeling scheme developed by the authors. The developed finite element model of the fractured reservior exhibits two-phase structure consisting of rock matrix and fractures, which are practically discretized into triangular elements and zero-thickness elements, respectively. The approach demonstrates a favorable comparison with the results obtained from the thereotical and other numerical solution. Monte Carlo simulation technique is then employed to probabilistically analyze the evolution of coupled thermal–hydraulic behavior, which involves 1000 different realizations considering the uncertainty of fracture structure. The results illustrate that the coefficient of variation of the seepage field decreases from 0 to 0.29 at 1 a to 0.06 to 0.15 at 50 a, averaging at 0.07, and the peak coefficient of variation of the temperature field decreases from 0.61 to 0.52, averaging at 0.27. The uncertainty of the temperature field is varying against time and prominent around the proximity to the advancing cold front, upper and lower boundaries, primarily associated with variations in fractures and constraint of boundary conditions. The existence of highly permeable fractures leads to a pronounced heterogeneity in the seepage and temperature fields within enhanced geothermal system. Furthermore, the importance of taking the uncertainty of the fracture structure into account for geothermal reservoir evaluation is highlighted, to enhance the reliability of numerical modeling for geothermal process prediction.

Pseudo-transient flow path optimization of a solar tower receiver operating with supercritical carbon dioxide

Aiming to further promote the decarbonization of energy production processes, this article simulates and optimizes the pseudo-transient behavior of a cylindrical solar receiver subjected to a spatially varying concentrated heat flux. The model employed, which accounts for its thermal–hydraulic behavior, uses supercritical carbon dioxide (s-CO2) as the heat transfer fluid and is based on a literature-validated hybrid formulation that axially discretizes the tube of the receiver and simultaneously solves the 2-D temperature distribution within the cross-section of each node. The operational conditions imposed on the receiver were determined such that it can produce 10 MW in a recompression Brayton cycle. Also, the receiver’s surface was exposed to a spatially variable concentrated solar boundary condition based on the direct normal irradiance (DNI) of a representative day for each month of the year, which was selected from the typical meteorological year (TMY) in Seville (Spain). The results showed that the combined thermal–hydraulic efficiency of the receiver increases with the number of panels and that the mass flow flowing through each symmetrical side of the receiver should not necessarily be identical for achieving maximal efficiency. More importantly, the results showed that by simply changing the location of the s-CO2 inlet and outlet ports, one can increase the efficiency of the receiver by about 3 % while considering the exact same concentrated solar irradiation.

Investigation of thermo-hydraulics performance and vapour bubble behaviour over 5×3 smooth hybrid elliptical tube bundle under saturated cross flow boiling condition - An experimental study

In the current experimental study, the thermal-hydraulic behaviour of a novel smooth hybrid elliptical tube bundle (TB) is compared with the performance of the corresponding conventional smooth circular TB. The smooth hybrid elliptical tube, made up of AISI-304, is in the form of a conventional elliptical shape from the outside but a circular shape from the inside. The operating parameters encountered in the current investigation are Heat Flux (10–75 kW/m2), Mass Flux (19–100 kg/m2s), and pitch-to-diameter ratio (P/D) (1.25 - 1.95). The outcomes show that the thermal-hydraulics performance in the case of a smooth hybrid elliptical TB is higher than a conventional smooth circular TB. The augmentation in the heat transfer coefficient obtained with a smooth hybrid elliptical TB is limited to 12 % in the entire range. The pressure drop improvement is found to be 48 % to 120 % equated to the conventional smooth circular TB. However, the effective performance of the smooth hybrid elliptical TB measured in terms of H/P ratio (H/P = heat transfer coefficient/total pressure drop) is found to be in the range of 0.3–2.5 compared to the conventional smooth circular TB which is in the range of 0.2–1.1. This is evident from the present experimental study that the employment of the elliptical TB will lead to lower pumping power than a conventional smooth circular TB owing to its lower pressure because of the streamlined surface. The obtained two-phase heat transfer enhancement efficiency based on enhancement factor and pressure drop penalty factor showed a higher gap TB (P/D = 1.95) has higher performance. Moreover, for the first time, an effort has been made to analyze the physics associated with vapor bubble dynamics and interaction along the height of the smooth hybrid elliptical TB.

Development of the thermal hydraulic analysis code, TASS/SMR-S, under multi-dimensional motion for a reactor under inclination and rolling conditions

The development of a thermal hydraulic analysis code that considers ship motion is important because it would adjust for additional motion effects that are not present in a land-based system analysis code. The TASS/SMR-S code, a one-dimensional system analysis code, was modified to predict the thermal hydraulic behavior of a nuclear reactor under the conditions of an ocean environment. To validate the improved TASS/SMR-S code, the calculation results were compared with the analytical solutions for a conceptual problem involving swaying, rolling, and compounded motions; experimental data were collected under inclination and rolling conditions. The modified TASS/SMR-S code exhibited good agreement with analytical solutions for the conceptual problem and experimental results obtained under ship motion conditions. In conclusion, the modified code was able to simulate thermal hydraulic behavior under ship motion conditions.