Natural Circulation Flow Research Articles

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.

Numerical study on flow instability in parallel helical tubes under ocean conditions

The superimposed coupling of flow oscillations generated by transient external forces under ocean conditions and flow oscillations in parallel helical tubes during two-phase flow instability may lead to more complex flow oscillation phenomena in parallel helical tubes. However, there is still a lack of understanding regarding the effect of ocean conditions on the two-phase flow instability within parallel helical tubes. The two-phase flow instability in parallel helical tubes under ocean conditions is numerically investigated in the present study. The effects of ocean conditions on the flow behavior and instability threshold in parallel helical tubes are studied with mass flux 200 kg/m2s and inlet subcooling within 5–70 °C. The results show that the mass flow rate decreases as the inclination angle increases. A larger rolling amplitude or shorter period increases the flow and temperature oscillation amplitude. Two characteristic frequencies are found in the natural circulation flow instability under the rolling motion. As the rolling amplitude increases, the compound flow oscillation amplitude is larger and more chaotic with a certain phase shift. The effect of stationary inclining on the flow instability threshold is rather significant than rolling effect. The results of the present study can serve as a significant reference for the operation and design of the helical-coiled once-through steam generator under ocean conditions.

Numerical investigations on the performance of coupled NCL based passive heat removal system of an IPHWR

Passive heat removal systems (PHRS) are indispensable additional safety features in nuclear reactors to deal with station blackout (SBO) conditions. These systems are based on coupled natural circulation loops (C-NCL) connected in series configuration. Further, these system's design, operating conditions, working fluid, heat removal capacity, etc., are specific to the reactor type. In the present work, the operational performance of such first of a kind PHRS of 700 MWe Indian pressurized heavy water reactor (IPHWR) is investigated. For this purpose, a coupled two-phase NCL code is developed, validated and employed. This PHRS consists of two NCLs connected in HHVC-VHVC (horizontal heater and vertical cooler - vertical heater and vertical cooler) configuration. In this study, initially, the full power reactor thermal-hydraulic conditions are established, subsequently, the performance of the PHRS under SBO is investigated. The analyses demonstrated that the system is adequate to effectively eradicate the decay heat by establishing single and two-phase natural circulation (NC) flows of ∼4.5 % and ∼5.6 % of rated conditions in primary and secondary transport loops, respectively. The maximum PHRS capability is predicted to be ∼72 MWth which is more than 3 % of full power. The transients of flow, pressure, energy transfer and heat sink thermal conditions are analyzed when the system changes from forced to NC. Further, the influence of system conditions such as loop diameter, loop height, sink volume, sink temperature and heat exchanger area on PHRS performance was predicted and the changes in system parameters were analyzed. The vital parameters influencing the PHRS performance are identified based on sensitivity analyses. This investigation provides significant insight into the flow and heat transfer characteristics of two-phase and coupled NCLs in the PHRS.

Analysis of flow heat transfer characteristics of helium-xenon gas mixtures during natural circulation

Helium-xenon(He-Xe) gas mixture is a commonly used cooling medium in fourth-generation advanced reactor systems. The thermal conductivity of a passive waste heat system is related to the safety and reliability of the reactor system. Therefore, it is crucial to explore the passive characteristics of the helium-xenon gas mixture. The passive waste heat removal characteristics of a helium-xenon gas mixture are influenced by the flow and heat transfer characteristics of the mixture. However, existing research on the flow and heat transfer characteristics of helium-xenon gas mixture mainly focuses on single tube bundles and plate bundles with fixed flow rates and high Reynolds numbers. However, existing research on the flow and heat transfer characteristics of the He-Xe mixtures mainly focuses on single tube bundles and plate bundles with fixed flow rates and high Reynolds numbers, and most empirical formulas apply to Reynolds numbers greater than 10000. However, the natural circulation flow rate under the influence of gravity varies with the variation of heating power, and in most operating conditions, the Reynolds number is less than 10000. Due to the lack of research on the convective heat transfer characteristics of the He-Xe mixtures at low Re, this paper investigates the flow and heat transfer characteristics of the He-Xe under natural circulation conditions. A three-dimensional model of the natural circulation circuit was established using numerical calculation methods. Comparing the numerical simulation results of natural air circulation with empirical formulas, it was found that the error was less than 20 %. This proves that the SST k-ω turbulence model can be used to study the flow and heat transfer characteristics of He-Xe gas mixtures in natural circulation. On this basis, natural cycle numerical simulations of He-Xe mixtures with different mixing ratios were carried out at a heat flux of 5000 W/m2, as well as natural cycle numerical simulations of 40 g/mol He-Xe gas mixture at different heating powers, to study the effects of heating power and He-Xe mixing ratio on the natural cycle heat transfer characteristics of the He-Xe gas mixture. The existing He-Xe gas mixture flow heat transfer equation was modified, and the error of the modified Nu empirical equation was less than 12 %, which can be used for the development of the passive waste heat export system program for subsequent He-Xe gas-cooled fast reactors.

Experimental on Transient Heating and Cooling of Natural Circulation Flow using A FASSIP-02 Large Scale Experimental Facility

Experimental on Transient Heating and Cooling of Natural Circulation Flow using A FASSIP-02 Large Scale Experimental Facility

A review of natural circulation flow in integral effect test facilities

The proposed NCFM map was reviewed using natural circulation data identified in the PWR simulated integral effect test (IET) facility with a height (length) scale ratio of 1/1. Based on this, the characteristics of natural circulation flow in the ATLAS and FESTA facilities, which are major IET facilities in Korea, are compared. It was confirmed that the natural circulation flow characteristics were well maintained within the NCFM area even at the half-height test facility of ATLAS. It was also confirmed that the natural circulation flow characteristics were well maintained within the NCFM area at the full-height test facility of FESTA, and the natural circulation flow characteristics were also maintained outside the NCFM region for a relatively small amount of coolant inventories.

Assessment of MARS-KS prediction capability for natural circulation flow in passive heat removal system

Considering that system analysis codes are used for the evaluation of the performance of Passive Safety Systems (PSSs), it is important to investigate the capability of the system analysis code to reliably predict the heat transfer and natural circulation flow, which are the main phenomena governing the performance of a PSS. Since MARS-KS has been widely validated for heat transfer models, this study focuses on evaluating its capability to predict the single and two-phase pressure drops and natural circulation flow. The straight pipe simulation results indicate that the pressure drop predictions are reliable within ±5 % error margin for the single-phase flow and the errors of pressure drop up to −30 % for the two-phase flow. Through single-phase natural circulation flow analysis, it is concluded that the use of the appropriate K-factor modeling based on the flow regimes is important since the natural circulation flow rate in MARS-KS is mainly affected by the form loss factor modeling. With two-phase natural circulation flow analysis, this study emphasizes the behavior of the system could change significantly depending on the two-phase wall friction and pressure loss modeling. With the analysis results, modeling considerations for the PSS performance evaluation with the system analysis codes are proposed.

Multi-physics analysis of SBLOCA by coupled Neutronics/Thermal-Hydraulics full plant model

With the development of computational tools, the accident analysis would be performed by best estimate approaches. The employment of multi-physics codes would be considered as new calculation procedure to obtain high valuable results by improving calculations with specialized codes. The small break loss of coolant accident (SBLOCA) is an accident that has typically been analyzed using system codes with a limited number of Thermal-Hydraulics (TH) channels in the core region, neglecting the use of multi-physics calculations. This approach results in insufficient accuracy and resolution in both neutronics and TH calculations within the core zone. To address this issue, this study proposes the implementation of a Neutronics/Thermal-Hydraulics coupling with one-to-one mapping in the core zone, achieved through the development of a comprehensive plant TH model. This approach aims to enhance the accuracy and resolution of calculations, enabling the identification of local phenomena, parameters, and feedback during accident simulations, and the incorporation of these effects on the macroscopic cross-sections at each time step. The results of SBLOCA simulation reveals the effects of conservative/best estimate approach on the sequence of events, the initiation time of the safety systems, the critical controlled parameters of the power plant, and the precise understanding of the accident. Also, it was found that the radial power distribution becomes disrupted and loses its symmetry immediately after the SCRAM leading to different characteristics of steam generators (SGs) and natural circulation flow rate in each loop of primary side.

Design and construction of fluid-to-fluid scaled-down small modular reactor platform: As a testbed for the nuclear-based hydrogen production

This paper presents the construction results and design of the UNIST Reactor Innovation platform for small modular reactors as a versatile testbed for exploring innovative technologies. The platform uses simulant fluids to simulate the thermal-hydraulic behavior of a reference small modular reactor design, allowing for cost-effective design modifications. Scaling analysis results for single and two-phase natural circulation flows are outlined based on the three-level scaling methodology. The platform's capability to simulate natural circulation behavior was validated through performance calculations using the 1-D system thermal-hydraulic code-based calculation. The strategies for evaluating cutting-edge technologies, such as the integration of a solid oxide electrolysis cell for hydrogen production into a small modular reactor, are presented. To overcome experimental limitations, the hardware-in-the-loop technique is proposed as an alternative, enabling real-time simulation of physical phenomena that cannot be implemented within the experimental facility's hardware. Overall, the proposed versatile innovation platform is expected to provide valuable insights for advancing research in the field of small modular reactors and nuclear-based hydrogen production.

Analysis of the effect of thermal stratification on the natural circulation decay heat removal of sodium-cooled fast reactor

Thermal stratification is an important phenomenon in the process of natural circulation decay heat removal in sodium-cooled fast reactors (SFR). It affects the establishment of the natural circulation and the flow rate of the natural circulation in a station black out (SBO) accident, and therefore affects the safety of the reactor. In safety evaluation, programs are often required to perform analysis for various conditions and long-term transients. However, it is difficult to meet this requirement by adopting a multi-dimensional method or one-dimensional and multi-dimensional coupling method. Therefore, this paper proposes applying a thermal stratification model to the system analysis program THACS to analyze the thermal stratification phenomenon. First, the thermal stratification model is validated with experimental data. The validated model is applied to analyze a station black out accident in the SFR. The lumped parameter model and thermal stratification model are used respectively for the hot pool. By comparing the calculation results of the two models, the effect of thermal stratification of the hot pool on key parameters such as the natural circulation flow rate of the primary circuit, natural circulation flow rate in the inter-wrapper region, average core outlet temperature, maximum core fluid temperature, and the maximum temperature of cladding was analyzed, and the effect of thermal stratification on natural circulation decay heat removal of the SFR was evaluated. This work can provide a reference for the safety analysis of the reactor and contribute to developing a more conservative safety analysis and evaluation strategy.

Analysis of transient measurements with thermal feedback and coupled TRACE/PARCS calculations performed on the BME Training Reactor

This paper presents nuclear reactor transient measurements with thermal feedback and corresponding coupled neutronic–thermohydraulic simulations performed for the natural circulation cooled Training Reactor of Budapest University of Technology and Economics. During the experiments, the outlet temperatures of several fuel assemblies were measured with thermocouples, while the reactor power was simultaneously logged by an ex-core neutron detector. The numerical simulations were carried out with the internally coupled TRACE/Point Kinetics and TRACE/PARCS code systems. The parameterised group constant library of the PARCS neutronic model, as well as the reactivity coefficients of the Point Kinetic model, were generated by the Serpent 2 Monte Carlo code. In the former case, special attention was given to the applied diffusion coefficients. The 3D/1D TRACE calculation model was prepared with the SNAP graphical interface. It was shown that after applying the necessary corrections, related to the peculiar natural circulation flow and reactor parameters, the measurement and both simulation results showed good agreement in terms of peak nominal power and average assembly outlet temperature. Although several anomalies are presented and modelling details that need improvement are highlighted, this study suggests that the used code systems could be appropriate tools to reproduce and analyse transients performed in small-core, high-leakage, natural circulation cooled pool-type reactors.

Mechanisms and inhibition method investigation on condensation induced water hammer in an equal-height-difference passive heat removal system

The equal-height-difference passive heat removal system utilizes the ocean as an ultimate heat sink for residual heat removal, demonstrating obvious advantages for ocean nuclear power plant applications. However, the structure with a zero-height difference can trigger the reverse flow of subcooled water. This could cause serious condensation induced water hammer events. In this study, a fully visual experimental setup was established to investigate the occurrence features. First, two kinds of formation mechanisms were defined by the visual experimental results. In this regard, the subcooling and the isolated steam slug are the necessary conditions for the formation and development. In addition, the occurrence positions were found to be concentrated in the test pipe section near the water tank. Then, a butterfly check valve was designed to avoid the condensation induced water hammer events based on the formation mechanisms. The experimental results showed that the check valve usage can effectively inhibit the reverse flow of the subcooled water and thus achieve its purpose of eliminating condensation induced water hammer events. Moreover, the check valve transforms the two-phase natural circulation flow from the condensation oscillation mode to the stable boiling two-phase flow mode. As a result, it enhances the two-phase natural circulation capacity.

Natural convective non-Newtonian nanofluid flow in a wavy-shaped enclosure with a heated elliptic obstacle

A numerical investigation has been carried out in a wavy-shaped enclosure with an elliptical inner cylinder to find out the effect of an inclined magnetic field and a non-Newtonian nanofluid on fluid flow and heat transfer. Here, the dynamic viscosity and thermal conductivity of the nanofluid are also taken into account. These properties change with the temperature and nanoparticle volume fraction. The vertical walls of the enclosure are modeled through complex wavy geometries and are kept at a constant cold temperature. The inner elliptical cylinder is deemed to be heated and the horizontal walls are considered adiabatic. Temperature difference between the wavy walls and the hot cylinder leads to natural convective circulation flow inside the enclosure. The dimensionless set of the governing equations and associated boundary conditions are numerically simulated using the COMSOL Multiphysics software, which is based on finite element methods. Numerical analysis has been scrutinized for varying Rayleigh number (Ra), Hartmann number (Ha), magnetic field inclination angle (γ), rotation angle of the inner cylinder (ω), power-law index (n), and nanoparticle volume fraction (ϕ). The findings demonstrate that the solid volumetric concentration of nanoparticles diminishes the fluid movement at greater values of φ. The heat transfer rate decreases for larger nanoparticle volume fractions. The flow strength increases with an increasing Rayleigh number resulting in a best possible heat transfer. A higher Hartmann number diminishes the fluid flow but converse behavior is exhibited for magnetic field inclination angle (γ). The average Nusselt number (Nuavg) values are maximum for γ = 90°. The power-law index plays a significant role on the heat transfer rate, and results show that the shear-thinning liquid augments the average Nusselt number.

Transient characteristics analysis of residual heat removal system for Helium-Xenon mixture cooled small reactor system

Helium-Xenon (He-Xe) gas mixtures are characterized by stable physical properties, easy compression, and excellent thermodynamic performance, making them suitable for thermal cycle coolant for underwater reactors. Due to the unattended deep sea, and harsh and complex operating environment, the residual heat removal characteristics are directly related to reactors’ safety characteristics and stable and reliable operation ability. Therefore, studying the He-Xe gas mixture residual heat removal characteristics can provide an essential guarantee for reactor safety analysis. This article optimizes the physical property calculation model of the He-Xe gas mixture in the Brayton Cycle Reactor System Analysis Program (BRESA) and obtains a system property analysis program suitable for calculating the natural circulation of the He-Xe gas mixture. And a natural circulation system for the He-Xe mixed gas cooling reactor was constructed using the modified BRESA. Numerical calculations and theoretical analysis were conducted on the He-Xe gas mixture’s natural circulation flow process and residual heat removal characteristics. The study of the transient factors of temperature, flow rate, and other related parameters in the system circuit found that when the reactor is shut down, relying on the natural circulation system can effectively output the residual heat of the reactor core to ensure reactor safety. The relevant research results have important theoretical and scientific significance for the safety assessment and evaluation of the Brayton cycle reactor system.

Development of reduced-order thermal stratification model for upper plenum of a lead–bismuth fast reactor based on CFD

After an emergency shutdown of a lead-bismuth fast reactor, thermal stratification occurs in the upper Plenum, which negatively impacts the integrity of the reactor structure and the residual heat removal capacity of natural circulation flow. The research on thermal stratification of reactors has mainly been conducted using an experimental method, a system program, and computational fluid dynamics (CFD). However, the equipment required for the experimental method is expensive, accuracy of the system program is unpredictable, and resources and time required for the CFD approach are extensive. To overcome the defects of thermal stratification analysis, a high-precision full-order thermal stratification model based on CFD technology is prepared in this study. Furthermore, a reduced-order model has been developed by combining proper orthogonal decomposition (POD) with Galerkin projection. A comparative analysis of thermal stratification with the proposed full-order model reveals that the reduced-order thermal stratification model can well simulate the temperature distribution in the upper plenum and rapidly elucidate the thermal stratification interface characteristics during the lead-bismuth fast reactor accident. Overall, this study provides an analytical tool for determining the thermal stratification mechanism and reducing thermal stratification.

Preliminary code development for single-phase liquid natural circulation flow simulation based on two-phase compressible model

It is of great significance to study the thermal-hydraulic characteristics of Lead-based Fast Reactor. In these years, the computational fluid dynamics method has become an attractive way in thermal-hydraulic and safety simulations of liquid metal cooled reactors. In order to investigate thermal-hydraulic characteristics of single-phase natural circulation flow in liquid metal cooled reactors, a self-developed CFD code was developed in this paper. The new code was proposed based on the two-phase compressible model, including liquid and vapor governing equations and a series of auxiliary models. The commercial CFD software ANSYS FLUENT was used to verify and validate the new code by comparing their simulation results under a typical heat driven cavity flow case. The comparison showed that the results simulated by the new code had good agreement with the ones calculated by ANSYS FLUENT. Furthermore, by coupling the above CFD model with the fuel pin heat transfer model, the application was conducted based on a full-scale reactor steady-state simulation of a natural circulation driven LBE-cooled fast reactor. From the results, it can be concluded that the new code can simulate and evaluate thermal-hydraulic behaviors of the LBE-cooled reactor well.

Experimental investigation on characteristics of Multiple-Loops coupled natural circulation system

The passive safety design is advantageous because it simplifies the system and brings higher safety features. Passive residual heat removal system (PRHRS) is one of the most important passive safety systems, which is designed to cool down the reactor during accidental conditions. For floating nuclear power platforms, seawater could be utilized as the final heat sink of PRHRS through multiple-loops coupled natural circulation.To study the operating characteristics and thermal hydraulic behavior of multiple-loops coupled natural circulation system, the Natural Circulation Mechanism Experimental System (NCMES) has been constructed. The station blackout accident (SBO) scenario is simulated in NCMES, which leads to the stopping of circulation pump and operation of PRHRS. Based on the experimental data, an analysis was carried out using the thermal hydraulic system code RELAP5, aiming to validate applicability and accuracy, as well as predict the operating characteristics of the experiment facility. The comprehensive analysis of experimental and calculated results shows that when it comes to the SBO scenario, the releasing residual heat and sensible heat of fluid can be removed to seawater through triple-loops coupled natural circulation; the calculated results fit the experimental data quite well and the RELAP5 code is applicable to predict operating characteristics of multiple-loops coupled natural circulation and PRHRS. As a result of self-adaptive characteristic of natural circulation, with the heating power decreasing, the response rate of natural circulation flow rate and fluid temperature is slow.

Experimental and analytical study on the stability of a low aspect ratio single-phase natural circulation loop coupled to a parabolic trough collector

Use of passive heat transport systems in renewable energy eliminates operator and external power requirements, thereby reducing the cost. Hence, a horizontal heater and horizontal cooler rectangular natural circulation loop (NCL) of low L/D (110) and aspect ratio (0.54) is coupled to a parabolic trough in the present work. Experimental investigations of the loop's steady state and stability characteristics were carried out at various powers corresponding to the solar irradiances using water as the heat transfer fluid. SolTrace was used to determine the heat loads at different direct normal irradiance (DNI) for the selected loop and trough dimensions. The upper threshold for stability was experimentally observed for the first time and reported here. Stability analysis carried out for different loop diameters revealed that the power at which the upper threshold of instability occurs reduces significantly with increase in loop diameter, enabling its observation in the low-pressure low temperature loop employed in the present study. Stabilizing techniques like orifice insertion and loop inclinations were also studied in detail. The results showed stability at higher heat loads (> 350 DNI), while the instability at lower heat loads can be curbed using an orifice (β = 0.5) or a loop inclination (> 4°). The small inclination can eliminate multiple steady state solutions and hence stabilizes the loop. Besides, the reduction in natural circulation flow rate with inclination is much smaller compared to the orifice making its heat transport capability higher. Additionally, the steady state results were validated with 1-D analytical model giving a good match.

Experimental study on transient flow characteristics in an equal-height-difference passive heat removal system for ocean nuclear power plants

The equal-height-difference passive heat removal system (EHDPHRS) utilizes the ocean as the heat sink for infinite heat dissipation and has broad application prospects in ocean nuclear power plants. However, the equal height difference between the heat source and heat sink is susceptible to severe flow instability. In this study, a fully visualized experimental facility was constructed to analyze the transient flow characteristics in the EHDPHRS. The experimental results revealed three typical two-phase natural circulation flow modes in the EHDPHRS. The formation mechanism of each flow mode was analyzed using thermal–hydraulic parameters. The results revealed that the flow mode is determined by the steam–water direct contact condensation and reverse flow of subcooled water, and both factors promote each other. Moreover, condensation-induced water hammer (CIWH) was found to be the most severe flow instability in the EHDPHRS. CIWH events can lead to violent pressure peaks and reduce the natural circulation capacity. In addition, when the occurrence position of CIWH events is concentrated at the pipe outlet, the influence of CIWH events on the two-phase natural circulation flow capacity gradually decreases. These research results aid in the optimal design of the EHDPHRS and prevention and control of CIWH events.

Numerical Evaluation of Coolability Limits of External Reactor Vessel Cooling Using an Improved Thermal-Hydraulic System Analysis Code

The external reactor vessel cooling provides a means to stabilize the molten core inside the reactor vessel during the severe accident in nuclear power plants. For a credible evaluation of the chances of success of in-vessel melt retention, one requires high-fidelity predictions on the natural circulation flow around the reactor vessel and, in particular, the corresponding critical heat flux (CHF) on a hemispherical lower head. In this study, the coolability limit of external reactor vessel cooling in the Korean APR1400 reactor was investigated numerically by improving a thermal-hydraulic system analysis code, MARS-KS1.5. Since such system analysis codes did not incorporate a CHF model for a downward-facing hemisphere, three CHF correlations were newly implemented in the MARS-KS code, two of them taking into account the effect of the mass flux. A transient analysis on the CHF and surface heat flux on the lower head was carried out while the surrounding cooling water was heated to create the two-phase natural circulation flow. The heat transfer mode on the heated surface was used as a measure of evaluating the success of the strategy, and the simulation results were compared with those obtained via the default CHF model in the code. The sensitivity analysis on the heat load from the core melt was also performed.