Search

Highlights • RELAP5 is modified to predict DWO in parallel narrow rectangular channels. • Two-phase friction model and nucleate boiling model are the most influential models. • The modified results are in better agreement with experimental data. • The modifications are effective on prediction of threshold power and period. Abstract As one of the most common types of oscillatory instability, density wave oscillation (DWO) is of practical importance in the general analysis of two-phase boiling systems. Due to the confinement effect of narrow space, the two-phase phenomena in thin rectangular channels are probably different from those in conventional pipes with large diameters. As the widely accepted system analysis code, RELAP5 has the capability to simulate the DWO phenomena in narrow channels, but the accuracy is limited under the condition of different pressures, flow rates, or subcoolings. In this paper, modifications for related constitutive models are proposed based on single-variable sensitivity analysis. Two-phase friction model and nucleate boiling heat transfer model are selected as the most effective models to improve the predictability of RELAP5. The influence of both models mentioned above is analyzed and it reveals that they are related to two-phase pressure drop and the vapor void fraction distribution along the channel, respectively. The modified calculation is assessed against the experimental results under the wide range of pressure 1–10 MPa, and these comparisons are performed in the map of stability boundary with two dimensionless parameters, phase change number Npch and subcooling number Nsub. Then the heating power and the period of flow oscillation under threshold condition are assessed in detail. The results show that a better agreement can be obtained between the modified calculations and the measured experimental data. [ABSTRACT FROM AUTHOR]

Passive residual heat removal heat exchanger (PRHR HX) is a key equipment in advanced passive safety pressurized water reactors (PWRs), such as AP1000. Immerged in the In-containment refueling water storage tank (IRWST), the C-shape PRHR HX removes the residual heat using natural convection and boiling heat transfer during postulated accidents. Therefore, the safety operation of the PRHR HX is very important for the nuclear power plant (NPP). In this paper, three dimensional CFD simulation of the secondary side fluid flow and heat transfer of the PRHR HX is performed. The drift flux model is used to simulate the two phase flow phenomenon in the IRWST. The tube region is modeled by the porous media approach. The flow resistance in the tube region is calculated by empirical pressure drop correlation and two phase flow multiplier. The heat transfer rate from the primary side to the secondary side fluid is also evaluated by widely used heat transfer empirical correlations. Additional source terms corresponding to the flow resistance and removed heat in the tube region are added to the momentum and energy equation, respectively. The governing equations are solved by the commercial CFD package FLUENT. Three dimensional distributions of the fluid velocity, temperature and void fraction are obtained. The heat transfer characteristics of the tube bundle is analyzed. [ABSTRACT FROM AUTHOR]

Abstract: Steam generator (SG), as the primary-to-secondary heat exchanger and pressure boundary of primary loop, should be integrated and perform well in heat transfer ability. Flow characteristics of the secondary side fluid of SG are essential to analyze U-tube wastage caused by the flow-induced vibration and thermal stress. In this paper, secondary side two-phase flow was simulated based on the porous media model. Additional momentum and energy source terms were appended to the momentum and energy equations for porous media region, respectively. The additional momentum source contained the resistances of downcomer, tube bundle, support plate and separator. The additional energy source included the heat transfer from primary side to secondary side fluid. Solving the governing equations by ANSYS FLUENT solver yielded the distributions of velocity, temperature, pressure, density and quality, which can be used in the analysis of flow-induced vibration and separators. The thermal-hydraulic characteristics of hot side differed from these of cold side considerably. The minimum flow quality of cold side was 0.07, while the maximum one of hot side was 0.71; the average flow quality of outlet was 0.272. The flow rate in the gap of the hot side was 1.02 times of that of the cold side. [Copyright &y& Elsevier]

Abstract: In the present study, two-dimensional numerical simulation of single bubble dynamics during nucleate flow boiling has been performed using moving particle semi-implicit (MPS) method. A set of moving particles was used to represent the liquid phase. The bubble–liquid interface was set to be a free surface boundary which can be captured according to the motion and location of interfacial particles. The interfacial heat transfer rate was determined by the energy variety of interfacial particles. The bulk liquid velocities investigated ranged from 0.07 to 0.3m/s. The surface orientations varied from vertical to horizontal through 60°, 45° and 30°. Bulk liquid subcooling varied from 0 to 6.5°C and wall superheat from 2.0 to 20.0°C. The computational results show that the bulk liquid velocity and surface orientation influenced the bubble diameter and liftoff time. Bubble would slide along the heater surface before lifting off and the sliding velocity at liftoff increased with an increase in bulk liquid velocity. Bubble dynamic was related to bulk liquid subcooling as well as wall superheat. The numerical results have been compared with the experimental data. [ABSTRACT FROM AUTHOR]

Abstract: In the present study, single steam bubble condensation behaviors in subcooled water have been simulated using Moving Particle Semi-implicit (MPS) method. The liquid phase was modeled using moving particles and the two phase interface was set to be a movable boundary which can be tracked by the topological position of the interfacial particles. The interfacial heat transfer was determined according to the heat conduction through the interfacial liquid layer and the coupling between momentum and energy was specially treated. Computational results showed that the bubble experiences various deformations at lower degrees of liquid subcooling while it remains nearly spherical at higher degrees of liquid subcooling. The bubble lifetime is nearly proportional to bubble size and is prolonged at higher system pressures. Bubble lifetime obtained from the MPS method agrees well with the experiments of , however it is lower than the predictions of . The underestimation is caused by severe bubble deformation at lower degrees of subcooling. The present study exhibits some fundamental characteristics of single steam bubble condensation and is expected to be instructive for further applications of the MPS method to evaluate more complicated bubble dynamics problems. [Copyright &y& Elsevier]

Abstract: A thermohydraulic and safety analysis code-TSACC has been developed using Fortran90 language to evaluate the transient thermohydraulic behavior of the China advanced research reactor (CARR) under station blackout accident (SBA). For the development of TSACC, a series of corresponding mathematical and physical models were applied. Point reactor neutron kinetics model was adopted for solving the reactor power. All possible flow and heat transfer conditions under station blackout accident were considered and the optional correlations were supplied. The usual finite difference method was abandoned and the integral technique was adopted to evaluate the temperature field of the plate type fuel elements. A new simple and convenient equation was proposed for the resolution of the transient behaviors of the main pump instead of the complicated four-quadrant model. Gear method and Adams method were adopted alternately for a better solution to the stiff differential equations describing the dynamic behavior of the CARR. The computational result of TSACC showed the adequacy of the safety margin of CARR under SBA. For the purpose of Verification and Validation (V&V), the simulated results of TSACC were compared with those of RELAP5/MOD3 and a good agreement was obtained. The adoption of modular programming techniques enables TASCC to be applied to other reactors by easily modifying the corresponding function modules. [Copyright &y& Elsevier]

• Empirical correlations are fitted in the helical cruciform seven-rod bundle. • Mixed convection in the laminar flow region is analyzed. • Transverse mixing in inner, side, and corner subchannels is investigated. A novel type of nuclear fuel named helical cruciform fuel may realize self-supporting for adjacent elements and improve heat transfer, both of which contribute to increasing the power density of the reactor. The flow and heat transfer characteristics of a hexagonal helical cruciform seven-rod bundle were experimentally investigated in the present work. 200 sets of isothermal frictional resistance experiments and 199 sets of heat transfer experiments were finished. The laminar flow region (200< Re <1000), the transitional flow region (1000< Re <3000), and the turbulent flow region (3000< Re <16,000) were determined based on the experimental data. Several correlations are compared with the present experimental data. The helical cruciform single-rod correlation underestimates the Nusselt number in the seven-rod bundle for Re <7000, while it has an average relative error of 7.96 % with the resistance experimental data. The empirical correlations to calculate the frictional resistance coefficient and the Nusselt number were fitted. When the Richardson number is greater than 0.1 in the laminar flow region, the Reynolds number is less than 600, and flow is considered to be mixed convection. The temperature distributions of the inner, side, and corner subchannel are analyzed. The temperature of the inner subchannel is higher than the side and corner subchannel. The increase of Reynolds number is beneficial to the fluid transverse mixing between different subchannels. This work gives validation data to support the numerical model in the helical cruciform rod bundle. [ABSTRACT FROM AUTHOR]

• System level LBE thermal-hydraulics analysis code is developed. • Coupling effect of LBE oxidation corrosion and thermal-hydraulics is considered. • Influence of oxidation layer on heat transfer characteristics of LBE reactor is conducted. For Lead-Bismuth Eutectic (LBE) alloy cooled reactor, oxidation corrosion is very important and would affect the thermal hydraulic characteristics of the reactor. In this paper, based on LETHAC code developed by Xi'an Jiaotong University (XJTU) for system level thermal hydraulic analysis LBE reactor, a simulation module for oxidation corrosion is developed and added. The influence of oxidation layer on the heat transfer characteristics of the reactor is conducted. Based on LESMOR reactor designed by XJTU, a simulation calculation is carried out under the long-term operation conditions of the reactor considering the growth of oxidation layer. The distribution of the oxide layer in the reactor, the change of temperature appreciation caused by the oxide layer over time and the distribution in the reactor are obtained. The presence of oxide layer will affect the fluid temperature and wall temperature of reactor. The longer the run time, the greater the impact. This paper could provide solid analysis methodology reference for the design of LBE reactor. [ABSTRACT FROM AUTHOR]

• The HCOTSG module is implemented in RELAP5. • The separate-effect validation for pressure drop and heat transfer is performed. • The integral calculation is performed for the HCOTSGs of IRIS and MRX. • The code can be used to analyze the thermal-hydraulic behavior of HCOTSGs. Helical coil once-through steam generators (HCOTSGs) have been widely used in the design of small modular reactor (SMR) during the past several decades. As the widely accepted system analysis code, RELAP5 underestimates the heat transfer capability and pressure drop of helical coil tube steam generators using the original geometrical parameters of HCOTSG, because the built-in thermal-hydraulic empirical correlations are only suitable for straight tubes. In this study, thermal-hydraulic models for helical coil tubes and tube bundles are implemented in RELAP5 while the original functions of the code are not affected. A new flag for helical component is proposed and heat transfer boundaries for tube side and shell side are developed. The separate-effect validations of friction factor and heat transfer in helical coil tube are carried out based on experimental data and code-to-code verification. Then, the original RELAP5 and developed RELAP5-HCOTSG are used to simulate helical coil tube steam generators of two SMRs. One is the integral reactor IRIS, and the other is MRX. The calculated results are compared to the design parameters. It shows that the HCOTSG module developed in this study can improve the capability of RELAP5 to predict the thermal-hydraulic characteristics in SMR once-through steam generators with helical coil tubes. [ABSTRACT FROM AUTHOR]

• A design analysis code for sodium heated OTSG is developed based on the power mesh method. • The code is verified with various design data both of integral type and modular type OTSGs. • The influence of tube diameter on heat transfer capacity has been analyzed by the code. Sodium heated once-through steam generator (OTSG) is a critical component of Sodium cooled Fast Reactor (SFR). It is a counter current heat exchanger. It transfers heat from hot sodium to water. Therefore, the steam generator is a necessity for the operation of the plant. However, due to the phase change exists in water side, it becomes challenging to design a steam generator for SFR. After the successful construction of CEFR (China Experiment Fast Reactor), a power-up demonstration SFR is being developed in China. For the design and analysis of the steam generator of the demonstration SFR, a thermal hydraulic Design Code of Once-through Steam generator for Sodium cooled fast reactor (DeCOSS) has been developed based on the movable mesh method. The DeCOSS code has been benchmarked based on the design data of OTSGs reported. The tube lengths and temperature distributions calculated by DeCOSS showed general agreements with published data. The influences of tube diameters on heat transfer capacity have been analyzed. The results indicated that the heat transfer capacity is determined by the thermal resistance of tube wall under the condition of fixed heat transfer coefficient of water side. In addition, under the same flow area of water and sodium, the heat transfer capacity could be enhanced by decreasing the tube diameters with constant tube wall thickness. Hence, the code can be utilized to do the sensitivity analyses and optimum thermal design of OTSGs. Moreover, it can support the design work of once-through steam generator of SFR. [ABSTRACT FROM AUTHOR]

Highlights • A one-dimensional code for thermal-hydraulic analysis of sodium heated OTSG is developed. • Both integral type and modular type sodium heated OTSG can be analyzed by the code. • The accuracy of the code is validated with various design data and the ETEC shutdown experiment. Abstract Sodium heated once-through steam generator (OTSG) is an essential component of Sodium cooled Fast Reactor (SFR). It transfers the heat from hot sodium to water and converts water to superheat steam. Steam can roll the turbine and generate electricity. Therefore, the safety of the steam generator is vital to the operation of the plant. China is developing the China Demonstration SFR (CDSFR) after successful construction of China Experiment Fast Reactor (CEFR). In order to make sure the safety of the steam generator under various transient conditions and support the design of the steam generator for CDSFR, a Transient analysis Code of Once-through Steam generator for Sodium cooled fast reactor (TCOSS) has been developed. To benchmark the developed physical model, the calculated results of the code have been compared with some design data of OTSGs and they are found in a good agreement. In addition, the transient results of the code have been compared with the experimental results of a shutdown experiment for the validation of the code. The predictions of the code agree well with the experimental data. Hence, the code can be utilized to predict the variations of thermal-hydraulic parameters in steam generator under different transient conditions. In addition, predictions can support the design of the steam generator for CDSFR. [ABSTRACT FROM AUTHOR]

Highlights • Heat transfer experiments for super-low finned-tube bundles in MSRs are performed. • Tests are carried out under the operating conditions of the nuclear power plant. • A more accurate heat transfer correlation under wider Re number range is proposed. Abstract Experimental investigation on the heat transfer characteristics of superheat vapor through a new designed moisture separator reheater with "super-low" finned-tube bundles for Generation III+ nuclear power plant, was carried out based on a 5*30 tube bundle test section. Impacts of the vapor mass flow rates are analyzed on the heat transfer characteristics of superheat vapor through the finned-tube bundle. A group of experiments are reviewed on single-phase fluid flowing across the triangular super-low finned-tube bundles whose fin heights are less than 2.75 mm and ratios of fin height to fin diameter are from 0.05 to 0.15. Combining the presented experimental data with the other open published data for single-phase turbulent gases flowing across triangular super-low finned-tube, a new heat transfer correlation is proposed for the MSRs super-low finned-tube bundles, which shows better accuracy and has wider Reynolds number range in application. [ABSTRACT FROM AUTHOR]

• An innovative three-dimensional thermal–hydraulic code, CorTAF, was developed based on OpenFOAM. • The CorTAF is more computationally efficient with the spatial resolution of subchannel level. • The CorTAF is validated against experiment benchmarks and subchannel analysis codes. • The steady state and transient simulations of a full scale PWR core were carried out using CorTAF. In this paper, a thermal hydraulic analysis method based on open source CFD platform OpenFOAM was proposed. The models of coolant flow and heat transfer, fuel rod heat conduction and coupled heat transfer were established according to the bundle structure characteristics of PWR core, and then a nuclear reactor three-dimensional thermal–hydraulic characteristics analysis code CorTAF was developed based on finite volume method with the spatial resolution of subchannel level. The international benchmarks, including GE3 × 3, Weiss and PNL2 × 6 fuel assembly flow and heat transfer experiments, were selected to carry out the validation. The calculation results were basically consistent with the experimental data, illustrating that CorTAF is suitable for predicting the flow and heat transfer characteristics of coolant in the rod bundle fuel assembly. The thermal–hydraulic characteristics of fuel assembly and PWR core under full power operation and blockage accident conditions were simulated using CorTAF. The steady-state and transient spatial distribution of physical quantities in subchannel-scale including coolant and fuel rod temperature were obtained, and the influence of blockage condition on flow and heat transfer characteristics was analyzed. This work has reference significance for the development of core thermal hydraulic analysis tools for PWR. [ABSTRACT FROM AUTHOR]

• Flow and heat transfer experiments of He-Xe gas mixture is conducted. • Friction factors and Nu numbers for 4000< Re <3 × 104 are obtained. • Empirical correlations are compared and recommended. The helium gas mixed with xenon gas is increasingly adopted as the working fluid for closed gas turbine cycles to optimize power/weight ratio and system efficiency. The flow and heat transfer performance of the He-Xe gas mixture has received widespread attention. In this paper, the closed gas loop with vertical circular test section is built, and the convective heat transfer characteristics of He-Xe gas mixture is experimentally investigated during the Reynolds number range of 4000–3 × 104. The wall temperature and gas pressure are measured, and the correlations of Nusselt number and friction factor versus Reynolds number are obtained. The main findings are as follows: the overall average friction factor can be predicted by the Blasius correlation and Taylor's friction correlation; the local Nusselt number agrees well with the Pickett correlation for Re >8 × 103; new correlations are proposed to predict total Nusselt number for 104 < Re <3 × 104 and 4000< Re <104, respectively. This study may contribute to the thermal design and optimization of heat exchangers concerning He-Xe gas mixture. [ABSTRACT FROM AUTHOR]

Porous media model in Fluent code, coupled with two-phase mixture flow model, resistance model of tubes and heat transfer model through tubes, is employed to develop a steam generator thermohydraulics analysis code STAF (Steam generator Thermohydraulics Analysis code based on Fluent). In this code, the heat transfer from primary to secondary side is calculated three-dimensionally during iteration. The localized velocity, temperature, enthalpy, quality and void fraction in steam generator can be obtained by this code. STAF is validated in two ways. First, STAF is used to calculate the thermal-hydraulic parameters in steam generator of AP 1000. The calculated results are compared with designed values to prove that the coupled heat transfer calculation in STAF is accurate. Second, STAF is employed to simulate the FRIGG test to validate the localized parameter calculation performance by comparing the calculated localized void fraction with test values. [ABSTRACT FROM AUTHOR]

In the present study, we develop a Steam-generator Thermohydraulics Analysis code based on Fluent (STAF) for predicting the three-dimensional localized thermal–hydraulic characteristics in the primary and secondary sides of steam generator. STAF code is developed based on the porous media model in Fluent. The flow resistances caused by the tubes, support plates, downcomer and separators are introduced to the momentum equation as additional source terms of shell side fluid; the heat transfer from primary to secondary side fluid is considered as the source term of energy equation of secondary side fluid. The flow and heat transfer in primary side, as well as the tube-to-shell-side heat transfer are solved by the user-defined functions in Fluent. STAF is used to simulate the Westinghouse MB2 test, and localized thermohydraulics parameters are obtained. The numerical results show good agreement with experimental results, demonstrating the ability of STAF to model the three-dimensional flow and heat transfer characteristics in primary and secondary side of steam generator. Besides, parameters associated with flow-induced vibration are also analyzed. [ABSTRACT FROM AUTHOR]

The relocation mechanisms of molten core have an important effect on the core degradation process which is a complicated coupled process of oxidation, flow, and heat transfer with great uncertainty. However, most relocation experiments are complicated integral-effects bundle experiments, too many factors are involved, and it is difficult to obtain the flow resistance correlation for the molten pool flow in the bundle channel. So in this paper, a separate-effects visualization experiment study on the flow mechanism of the molten pool relocation was conducted by using glycerol-water solutions at room temperature with a 2 × 2 bundle channel. The flow front of glycerol-water solutions can be captured by the high-speed camera via a transparent plexiglass shroud. And a new flow resistance correlation of the bundle channel for the molten pool relocation is obtained. In addition, the influence of flowing gas on the relocation process of the molten pool is also investigated. This study is supposed to provide valuable references for the calculation model of the molten pool relocation in severe accident analysis codes. • A separate-effects visualization experiment study on the flow mechanism of the molten pool relocation was conducted in room temperature. • A new flow resistance correlation of the bundle channel for the molten pool relocation is obtained. • The influence of flowing gas on the relocation process of the molten pool is investigated. [ABSTRACT FROM AUTHOR]

Spray cooling is an efficient cooling technology with high heat dissipation, and has been extensively utilized in various industry areas. The interaction mechanism between the spray droplet and environmental space with high pressure and temperature is extremely complex, especially with non-condensable gas. In this paper, a pseudopotential multicomponent multiphase lattice Boltzmann model for droplet migration with non-condensable gas in high temperature and pressure was proposed. The mechanism of transfer characteristics for droplet in steam were studied. It is found that the mass transfer of droplet is controlled by density and condensation, droplet growth is mainly impacted by density at initial stage, and in later period, the growth is greatly weakened by the condensation suppression of non-condensable gas. The calculated results show that when Fo > 0.4, mass transfer is significantly decreased. Heat transfer is also affected by non-condensable gas, and 2% gas declines the droplet heat transfer rate by about 10%. It is also found that the mass transfer rate is approximately linearly associated with the ratio of non-condensable gas content inside and outside the droplet. The results can be used for development of spray cooling model, and have positive guiding significance for optimizing spray cooling system. • A heat and mass transfer model of droplet and steam with non-condensable is established. • The mechanism of transfer characteristics for droplet in steam were studied. • The mass transfer of droplet is controlled by density and condensation in different stage. • The mass transfer rate is approximately linearly associated with the ratio of non-condensable gas content. [ABSTRACT FROM AUTHOR]

The thermohydraulics of the steam generator secondary is analyzed using Fluent's porous media model coupled to a two-phase flow mixture model. The resistances introduced by the tubes, supports, downcomers and separators are calculated as the source terms of the momentum equation. The heat transfer from primary side fluid to the secondary is calculated three-dimensionally each iteration and is supplied as a heat source on the secondary flow field calculation. The effects of the power level on thermal hydraulic characteristics in the steam generator are also investigated. Flow-induced vibration damage increased in severity with increasing power level. [ABSTRACT FROM AUTHOR]

Eulerian two-fluid model coupled with wall boiling model was employed to calculate the three dimensional flow field and heat transfer characteristics in a hot channel with vaned spacer grid in PWR. The heat transfer from pellet-gap-cladding to coolant was also taken into account by a system coupled code MpCCI. The wall boiling model utilized in this study was validated by Bartolomei experiment data, and a good agreement can be observed. By solving the governing equation in a two-way coupled method, the distribution of temperature in the pellet-gap-cladding region and the distribution of temperature, void fraction and velocity of two-phase flow in coolant channel can be obtained. The influences of spacer grid and mixing vane on the thermal-hydraulic characteristics were analyzed. The heat transfer capacity was strongly improved by the spacer grid and mixing vane, while the flow resistance was also enlarged. Localized volume fraction of vapor phase decreased due to mixing vane, which will decrease the possibility of the departure from nucleate boiling (DNB) and increase the critical heat flux (CHF). By analyzing the temperature and void fraction at cladding outer surface, the critical regions where hot spot may occur were determined. [ABSTRACT FROM AUTHOR]

• SACOS-PLATE: Enhanced code for plate fuel analysis. • Validated against RA-6, Bettis experiments. • Unique heat conduction, flow models. Over the past few decades, plate-type fuel elements have been widely used in advanced research reactors, small modular pressurized water reactors, production reactors, etc., with the advantages of compact structure arrangement, good heat transfer characteristics and high fuel consumption. Currently, a thermal–hydraulic subchannel analysis code, namely SACOS-PLATE code, was newly designed to satisfy the calculation of different types of plate fuel elements under various steady-state operating conditions. Based on the original SACOS code, this updated version modifies the flow distribution model, the plate heat conduction model, the heat distribution model as well as the flow heat transfer model, and then realizes the simulation of the special thermal–hydraulic phenomena of the fuel plate and the rectangular flow channel. In this present study, the RA-6 and Bettis experimental data was used to validate the code, and the simulation calculation was performed on the JRR-3 research reactor, which effectively proved the good computing capability of SACOS-PLATE code. [ABSTRACT FROM AUTHOR]

The lead-bismuth cooled fast reactor core adopts box-type design, and the fuel rod bundle is fixed by wrapping wires. The complex geometric structure of the fuel assemblies significantly affects the flow and heat transfer status of the liquid metal coolant. Building upon the open-source CFD platform, OpenFOAM, this work established conservation equations for liquid metal coolant, fuel rod thermal characteristic models, inter-wrapper coolant analysis models, and coupled heat transfer models, and developed the subchannel level thermal-hydraulic analysis code CorTAF-LBE for whole-core analysis of LBR core using the finite volume method. The wire-wrapped rod bundles internal flow and heat transfer experiments of lead-bismuth, carried out based on the THEADES and KYLIN-II circulation loops, were selected to perform verification and validation. The results match well with the experimental data and CFD fine-modeling prediction. The steady-state numerical simulation of three wire-wrapped rod bundle assemblies with inter-wrapper flow was carried out under normal operation condition. The temperature and pressure distributions of the coolant in each subchannel and inter-wrapper flow region, the temperature of cladding and other key thermal-hydraulic parameters were obtained. The role of the inter-wrapper coolant in heat removal from the fuel assemblies was revealed. The work in this paper has referential value for the optimization of lead-bismuth cooled fast reactor design and safety assessment during operation. [ABSTRACT FROM AUTHOR]

Pressurizer is widely used in the industrial systems, which can maintain the pressure of the system in a certain range. Especially, as the system pressure rises, the spray droplets interacts with the saturated steam in pressurizer to reduce system pressure. In this paper, a heat and mass transfer model of droplet and steam in pressurizer is established through the lattice Boltzmann method (LBM), and the characteristics of moving droplet in steam space with different Reynolds numbers are analyzed. In this paper, these results show that the flow fields on both sides of the droplet have obvious symmetry at low Reynolds number. As the Reynolds number increasing, the wake region of flow field for moving droplets gradually loses symmetry. When the Reynolds number reaches 103 orders of magnitude, the droplet motion begins to appear horizontal deviation. The growth of droplet condensation radius is inhibited with the increase of Reynolds number. When Reynolds number reaches 2600 and Fo = 0.2, the growth of droplet radius stops and then decreases. Due to the influence of inertia force generated by internal circulation flow and droplet movement, as the Reynolds number increasing, the droplet heat transfer rate decreases at first, but begins increasing when Reynolds number raises to about 2000. The calculation results have reference value for the development of spray droplet model of pressurizer. • A heat and mass transfer model of droplet and steam in pressurizer is established. • The characteristics of moving droplet in steam under different Re are analyzed. • With the increase of Re , the growth of droplet condensation is inhibited. • Heat transfer rate decreases first and then increases with the increase of Re. [ABSTRACT FROM AUTHOR]

• Sliding mesh method was used to simulate moving conditions. • Comparing the numerical results with the experimental results. • Two calculation methods of simulating moving conditions were compared. • It is concluded that the effect of rolling motion is the most sever. • Secondary flows and flow instabilities under moving conditions were analyzed. The investigation of flow and heat transfer of lead-bismuth eutectic (LBE) under moving conditions is vital to the design optimization of LBE loop system in offshore engineering. In this paper, we have developed a method for simulating moving conditions using sliding mesh. Moreover, we have conducted experiments to validate the numerical model, and the comparison results demonstrate a good agreement between them. The calculation results demonstrate that the period of the flow rate fluctuation is consistent with the motion period, and the amplitude of the fluctuation will increase with the increase of the maximum range of motion and the decrease of the motion period. In addition, the influence of rolling on system parameters is more severe than that of heaving. These results obtained in this study have important engineering application value for the design of LBE loop system under moving conditions. [ABSTRACT FROM AUTHOR]

Abstract: In this study, the 3D flow and heat transfer characteristics in rod bundle channels of the super critical water-cooled reactor were numerically investigated using CFX codes. Different turbulent models were evaluated and the flow and heat transfer characteristics in different typical channels were obtained. The effect of pitch-to-diameter ratio (P/D) on the distributions of surface temperature and heat transfer coefficient (HTC) was analysed. For typical quadrilateral channel, it was found that HTC increases with P/D first and then decreases significantly when P/D is <1.4. There exists a “flat region” at the maximum value when P/D is 1.4. If P/D is larger than 1.4, heat transfer deterioration (HTD) occurs as main stream enthalpy is quite small. Furthermore, the HTD under low mass flow rate and the non-uniformity of circumferential temperature were also discussed. [ABSTRACT FROM AUTHOR]

Summary: Heat pipe cooled reactor is designed to meet the future energy demands with many unique advantages such as modularization, solid and static state, high reliability and passive safety. In the innovative nuclear reactor design, heat pipes are used to cool the reactor core and transfer the fission heat passively, thermoelectric generators (TEGs) are utilized to achieve static energy conversion. In this article, a finite element model is established to investigate the detailed TEG thermoelectric performance coupled with heat pipes. A temperature field and flow field coupling model is used to simulate the heat transfer performance of the potassium heat pipe including a heat pipe wall, wick and vapor region. A heat pipe TEG experimental setup is established and the steady‐state experiment is conducted. The experimental setup is selected as the numerical simulation object and the simulated results are compared with the experimental data to verify the numerical model. The maximum simulation error is within 5%. Based on the verification and validation, the steady‐state thermoelectric performances, the effect of contact thermal resistance and external load on the skutterudite TEG thermoelectric performances are investigated. The detailed heat flux distributions, temperature distributions and potential distributions are researched in the steady‐state simulation. The temperature distributions of the TEG cross section are centrally symmetrical and the horizontal temperature difference is 26°C. The temperature difference introduced by thermal resistance at the TEG hot side increases from 33°C to 94°C with the increasing heat flux through TEG module. The influence of thermal resistance on TEG hot side is more significant than that on TEG cold side. The effect of external load on TEG output characteristics is investigated under different heat flux conditions. With the increase of heat flux from 60 to 166 kW/m2, the temperature difference of TEG module and p‐n unit increase by 432°C and 348°C, the thermoelectric conversion efficiency is improved from 3.6% to 8.5%, and output power increases by 20 W, respectively. The research results are of guiding significance for exploring the application potential of TEGs in the static heat pipe cooled reactor. This work provides research foundations for the multiphysics coupling analysis of the micro heat pipe cooled reactor. [ABSTRACT FROM AUTHOR]

Summary: Closed type assembly leads to complicated core thermohydraulics which are significant for design and safety analysis in sodium‐cooled fast reactor (SFR). Most of the major phenomena can be dealt with by general parallel channel network and two‐dimensional model for inter‐wrapper flow (IWF) in system code transient thermal‐hydraulic code for analysis of SFR (THACS). A main deficiency left is the overestimation of inter‐assembly heat transfer effect due to lump temperature message that one‐dimensional assembly model can only provide. A methodology was thus developed to compute wall‐subchannel temperature for simulating energy transportation through duct wall. Intra‐assembly thermohydraulics were classified as two mechanisms and non‐dimensional parameter was introduced for characterizing each of them. Ratio between linear heat flux density through hexagonal wrapper surface and pin surface qw/p was employed to account for radial heat transfer. Flow redistribution effect arouse in mixed and natural convection was concretized by a importance symbol of buoyancy force, Richardson number Ri. Two criteria were also proposed for dividing flow state into three parts and specific model for each section was defined. Two validation cases against 3‐D computational fluid dynamics (CFD) simulation of China experimental fast reactor whole core and test data from plant dynamics test loop with direct heat exchanger experiment were performed with THACS. Noticeable improvement after incorporation of this model demonstrated its value in system analysis. Although the procedure for parameter computations is geometry related, this method can be reproduced to any type of assembly directly. [ABSTRACT FROM AUTHOR]

Summary: In this paper, development and preliminary validation of an in‐house code LETHAC (lead‐based fast reactor thermal‐hydraulic analysis code) is introduced. LETHAC is developed for predicting the thermal‐hydraulic transient performance of a lead‐bismuth reactor. The mathematical models and module relations of LETHAC are presented in detail. The most‐used LBE (lead‐bismuth eutectic) flow and heat transfer correlations are compared against data from KYLIN‐II 61‐rod wire‐wrapped bundle experiments. Results show little difference between the chosen heat transfer correlations as the relative error of these correlations is all within 2%. Novendstern correlation is recommended for wire‐wrapped bundle frictional resistance coefficient calculation as it fits the experimental data best. Besides, the transient calculation capability of LETHAC is also preliminarily evidenced by TALL transient tests. Good agreements are shown between the calculated temperatures and the experiment ones. Relative errors between calculation results and experimental data are almost within 10% during the whole transient process, which indicates LETHAC can be used for predicting the transient behavior of lead/lead‐bismuth cooled systems. [ABSTRACT FROM AUTHOR]

Summary: The fuel rods in the lead‐based fast reactor (LFR) are usually supported by wire spacer, and the presence of the wire results in the complexity of the thermal‐hydraulic analysis in rod bundles. In this study, several previous experimental research about lead‐bismuth eutectic flowing through the rod bundle are investigated. Based on the collected experimental data, a series of flow and heat transfer correlations in the wire‐fixed rod bundle are compared and evaluated. The recommended correlations are then implemented into the in‐house sub‐channel analysis code SACOS‐PB. Furthermore, to account for the changes in the channel cross‐section parameters caused by the wire spacer, two optional wire geometric treatments were added to the code: axially‐averaged approach and axially‐varied approach with extra cross‐flow enhancement model. 19‐rod bundle and 61‐rod bundle tests are used to validate the optimized code. The deviations for the temperature differences of cladding and coolant in internal sub‐channels are within 25%, and the axially‐varied approach shows a higher prediction accuracy, which proves that the improvement of the SACOS‐PB are reliable. This work could provide a reference for the subsequent design and development of the LFR core. [ABSTRACT FROM AUTHOR]

Summary: Helium‐Xenon (He‐Xe) gas mixture featured with low Prandtl number is increasingly used in various energy devices as heat‐carrying agent. To figure out the heat transfer characteristics of He‐Xe gas mixture in uniformly heated circular and tri‐lobe channels, numerical investigations are conducted and presented in the article. The adopted calculation models are validated by comparison between simulation results and experimental data. The local and total heat transfer characteristics are obtained, and new temperature correction factors for the circular and tri‐lobe channels are proposed and recommended. The simulation results obtained by using the SST k − ω turbulent model agree well with the experimental data, this kind of turbulent model matched with appropriate turbulent Prandtl number model could provide satisfactory temperature predictions for He‐Xe gas mixture in a wide range of temperature ratio Tw/Tb. The simplified Petukhov correlation with temperature correction can give conservative predictions with relative error less than 10%. For the tri‐lobe channel, due to the local restricted flow, the gas temperature at the corners is much higher than that at the concavities. The local wall temperature and local Nu number vary azimuthally, showing a large temperature gradient. The tri‐lobe channel has better heat transfer performance than that of the circular tube when the temperature ratio Tw/Tb is no more than 1.5, but it deteriorates significantly with the increasing temperature ratio. This article may contribute to the numerical simulation and thermal design of heat transfer components concerning He‐Xe gas mixture. [ABSTRACT FROM AUTHOR]

The strong coupling heat transfer between subassemblies and inter-wrapper gaps is closely related to the safety of the sodium-cooled fast reactor (SFR). Therefore, it is very important to carry out a detailed thermal hydraulic analysis of the core. In this paper, the inter-wrapper channel is modeled using the subchannel analysis method. A core thermal-hydraulic analysis code SACOS-Na (Subchannel Analysis Code Of Safety for SFR) is developed by using SIMPLE algorithm. For the subassembly module, a total of three experiments, including ORNL 19-pin experiment, TOSHIBA 37-pin experiment and KNS 37-pin experiment, are adopted to demonstrate the feasibility of the mathematical and physical models and numerical algorithms of the SACOS-Na as well as the rationality of the selected auxiliary models. For the inter-wrapper gap module, WEISS 2-subassembly experiment and CCTL-CFR 3-subassembly experiment are used to verify not only the simulation capability for backflow and strong inter-subassembly heat transfer, but also the heat structure model and the subchannel modeling approach on inter-wrapper gaps. In addition, the prediction accuracy of SACOS-Na has reached the advanced international level by comparing with the worldwide well-known core thermal hydraulic analysis codes. • The subchannel modeling method is adopted to build the mesh for inter-wrapper channels. • The SIMPLE algorithm is used to solve the problems of the reverse flow and inter-subassembly heat transfer occurred in SFRs. • The universal core thermal-hydraulic analysis code SACOS-Na for SFRs is developed. • The accuracy and reliability of the SACOS-Na are discussed from all aspects. [ABSTRACT FROM AUTHOR]

Helical coil once-through tube steam generator (HCOTSG) is an excellent form of heat exchanger. HCOTSG is adopted in various industries, such as Lead-bismuth Fast Reactor (LFR). The flow and heat transfer in the shell side of HCOTSG are important for its design and optimization. It is of great significance to carry out relevant experiments to obtain empirical correlations. In this paper, experimental study on the lead-bismuth eutectic (LBE) cross flow around the tube bundle is carried out to investigate the thermal-hydraulic characteristics based on geometric approximation of local region. Three test sections with different inclination angles of 0°, 4°, 8° are adopted in the experiments with Peclect number (Pe) up to 350. The experimental sections are uniformly heated with heat flux ranging from 20 to 100 kW/m2. The heat transfer characteristic of LBE cross flow around inclined tube bundle is obtained, considering the effects of inclination angle of the test section. The heat transfer correlation of LBE cross flow around inclined tube bundle is proposed for the first time. By comparing the data of the experiments, the results are all within the error range of 10%. The correlation is appropriate for the program development of LBE HCOTSG. This study can provide essential experimental data for the design and code development of LBE HCOTSG. [ABSTRACT FROM AUTHOR]

• A thermo-mechanical coupling code based on annular fuel sub-channel code is developed. • The developed code has been verified by FROBA-ANNULAR. • The change of the gap heat transfer coefficient, the flow split and MDNBR has been discussed. To simulate the thermal hydraulic performance of the annular fuel more accurately, some mechanical models have been incorporated into the sub-channel code SACAF, including pellet displacement model, gap heat transfer model, fission gas release model and fuel-cladding mechanical interaction model. The results calculated by the developed code are in good agreement with that by FROBA-ANNULAR. The developed code can consider the change of the heat transfer coefficient of the inner and outer gaps, the release of the fission gas release, the inner pressure of the fuel, the stress and strain of the cladding, which cannot be calculated by the conventional sub-channel code. Based on the calculation results, a great change is found in the gap heat transfer coefficient and MDNBR during the operation, which implies that the coupling of the sub-channel code and the mechanical performance models is necessary. [ABSTRACT FROM AUTHOR]

Summary: Waste heat recovery helps reduce energy consumption, decreases carbon emissions, and enhances sustainable energy development. In China, energy‐intensive industries dominate the industrial sector and have significant potential for waste heat recovery. We propose a novel waste heat recovery system assisted by a heat pipe and thermoelectric generator (TEG) namely, heat pipe TEG (HPTEG)，to simultaneously recover waste heat and achieve electricity generation. Moreover, the HPTEG provides a good approach to bridging the mismatch between energy supply and demand. Based on the technical reserve on high‐temperature heat pipe manufacturing and TEG device integration, a laboratory‐scale HPTEG prototype was established to investigate the coupling performances of the heat pipes and TEGs. Static energy conversion and passive thermal transport were achieved with the assistance of skutterudite TEGs and potassium heat pipes. Based on the HPTEG prototype, the heat transfer and the thermoelectric conversion performances were investigated. Potassium heat pipes exhibited excellent heat transfer performance with 95% thermal efficiency. The isothermality of such a heat pipe was excellent, and the heat pipe temperature gradient was within 15°C. The TEG's thermoelectric conversion efficiency of 7.5% and HPTEG's prototype system thermoelectric conversion efficiency of 6.2% were achieved. When the TEG hot surface temperature reached 625°C, the maximum electrical output power of the TEG peaked at 183.2 W, and the open‐circuit voltage reached 42.2 V. The high performances of the HPTEG prototype demonstrated the potential of the HPTEG for use in engineering applications. [ABSTRACT FROM AUTHOR]

• Detailed heat transfer in blistered plates' coolant channels is modeled using CFD. • Multiple blisters can cause flow maldistribution beyond design threshold. • Single blister leads to a rather high fuel meat maximum temperature. • Multiple blisters' influence on fuel temperature is independent. Due to the special structure, the plate-type fuel element possesses a unique failure mode, i.e. blistering. Blistering is a kind of plastic deformation, induced by thermal stress and pressure of gaseous fission products at the interface of fuel meat and cladding. A blistering cladding surface will alter the flow channel shape and thus affect the flow and heat transfer characteristics in a fuel assembly. The influence is significant because the parallel channels are closed and narrow. In this study, three plates and channels were modeled to represent parallel channels in a fuel assembly. Two kinds of blisters were considered, namely round blister and pillow-like blister. Six cases with various blistering distribution were simulated to investigate its influence on pressure drop, mass flow distribution, and fuel temperature. The turbulent model and mesh condition were carefully chosen and validated. Detailed flow structures were visualized and their relations with flow and heat transfer characteristics were analyzed. Results show that multiple blisters can induce the flow distribution factor to exceed the design threshold value. A single blister, due to the large thermal resistance of gaseous fission products, can cause a relatively high maximum temperature in the fuel plate and lead to additional structure rupture. Multiple blisters' influence on fuel temperature is independent and similar with that of a single blister. Larger blisters lead to higher maximum fuel meat temperature and therefore are more dangerous. [ABSTRACT FROM AUTHOR]

• Melt pool temperature and wall heat flux were obtained within a two-layer corium pool. • Top cooling boundary condition had a significant impact on natural convection behavior. • Nu up /Nu dn provided valuable data for prototypical scenario with Ra′ of 1016. During core meltdown accidents, the heat flux distribution along the wall of lower plenum was closely related to the potential of vessel failure. Concerning with the natural convection heat transfer behavior in stratified corium pools, several two-layer tests were performed on COPRA facility recently. The test vessel (a 1/4 circular slice with an inner radius of 2.2 m and a width of 0.2 m) is divided into two parts by a plate made of 304 stainless steel. The lower part, heated by 18 electrical heating rods, is applied to simulate the oxide pool. The upper part without heat generation is assumed to be a metallic layer. During the experiment, different heating powers and boundary conditions were employed to obtain the temperature distributions of oxide pool and metal layer, as well as the heat flux through the vessel wall. Prototypic Ra′ range, varied from 1.88 × 1016 to 5.39 × 1016, was successfully achieved in the oxide pool. Despite the smaller downward Nu dn numbers, the upward Nu up numbers within this study were well in accordance with the reported data. It should be mentioned that the results of water are unrepresentative for the reactor situation since no crust was formed in the bottom and top of oxide pool. However, the current study will still attract interest by the qualitative analysis of top cooling influence on natural convection phenomenon and provide valuable information to other researchers who concern with the crust formation and focusing effect on stratified corium pools behavior. [ABSTRACT FROM AUTHOR]

• Two methods were proposed to study the effect of rolling. • Bubble rising were simulated, the results agree well with experimental data. • Influenced by the rolling motions, bubbles may deform more obviously. • The bubble tends to move laterally and periodically under rolling conditions. The accurate prediction of bubbles rising characteristics is very important for the two-phase flow and heat transfer calculation, especially under the ocean conditions. The additional force will be introduced and applied on the bubbles and the bubble behaviors is different from that under the stationary conditions. In present work, two methods, the additional force and dynamic mesh of Arbitrary Lagrangian–Eulerian (ALE) method, are compared, the result shows these two methods are in accordance to calculate the effect of rolling. Then Volume of fluid (VOF) and ALE method have been used to calculate the single bubble rising characteristic under rolling conditions. Results show that sphere bubble, ellipsoidal bubble and sphere cap bubble exist during the whole rising process and the deformation is becoming larger with the increase of bubble size. The bubble tends to move laterally and periodically under rolling conditions. The period of lateral movement is in accordance with the rolling period and the amplitude of the bubble is proportional to the rolling amplitude. Also, larger bubbles have the higher rising velocity than the small ones. The bubble rising characteristics are determined by not only the rolling motion but also the surface oscillation under rolling conditions. [ABSTRACT FROM AUTHOR]

The lead-bismuth eutectic(LBE)-cooled nuclear systems have significant natural circulation due to its large thermal expansion. Therefore, the passive safety of the reactor in postulated accidents is greatly improved. Moreover, the flow resistance and heat transfer characteristics are important to natural circulation ability of LBE-cooled system. In this paper, the natural circulation capability, the flow resistance and heat transfer characteristics were experimentally studied in the N atural C irculation C apability L oop(NCCL) facility. The research mainly includes four aspects: 1) start-up process of natural circulation; 2) capability of natural circulation; 3) flow resistance of LBE in an annular channel and 4) heat transfer in an annular channel. The results showed that a significant natural convection flow was generated in the experiment. Due to high thermal expansion of LBE, the natural circulation is quickly established and stabilized. The flow rate of natural circulation depends on the loop resistance and the temperature between heating section and cooling section. In our experiment, the maximum natural circulation flow rate was measured to be ∼0.81 kg/s. A new friction coefficient correlation was fitted and the results is slightly larger than the value calculated by the Blasius formula. For the convection heat transfer of liquid LBE in an annular channel, the heat conduction term is dominant, and the Nusselt number increases with the increase of Peclet number. Fitted experimental correlation of convection heat transfer of liquid LBE in annular channel is given. • The flow resistance and heat transfer experiments were conducted in a lead bismuth eutectic loop. • The start-up process of natural circulation and capability of natural circulation were studied in this experiment. • Fitted experimental correlations of friction coefficient and convection heat transfer of LBE in annular channel were given. [ABSTRACT FROM AUTHOR]

The thermal load heterogeneity will lead to inhomogeneous local heat flux and thermal stress, extremely detrimental to the operational stability of heat exchangers with multi-assembly structure. In this paper, a full-scale analysis method for the helical coil steam generator in industry was developed and implemented. The validation was performed against flow boiling heat transfer experiment of helical tube. The simulation of single assembly and full-scale helical coil steam generator in high temperature gas-cooled reactor was performed, and the distributions of thermal-hydraulic parameters in heat exchange region were obtained. The results show that the nonuniform helium flow distribution with the inhomogeneous factor D m value of 34.83% has a significant influence on the heat transfer characteristics. The peak heat transfer power of central assembly increases by 33.76% against design condition, introducing large thermal load heterogeneity with the inhomogeneous factor D p value of 30.45%, even threating the structural integrity of helical tube in long-term operation circumstances. The structural elements design in the superior compartment should be optimized to increase the helium mass flow rate of the assemblies below the nozzle and behind the helium baffle, improving flow distribution uniformity and reducing the peak value of local heat flux and thermal stress. • A full-scale analysis method of helical coil steam generator in HTGR was developed. • The flow distribution in multiple heat exchange assemblies was obtained. • The thermal load heterogeneity was quantificationally investigated. • The optimization suggestion of superior compartment elements was provided. [ABSTRACT FROM AUTHOR]

• Thermal-hydraulics of LBE is studied and an experiment is carried out. • A new empirical correlation of LBE is developed, and validated by XJTU 7-rod bundle, KIT 19-rod bundle, and ENEA 37-rod bundle. • Detailed CFD is carried out to analyze the characteristics of flow and heat transfer of LBE. • Turbulent heat transfer model of LBE is discussed and verified. Lead-bismuth eutectic (LBE) is a liquid with low Prandtl number and high thermal conductivity, generally used as the coolant in lead-cooled fast reactors of Generation-IV. The traditional Reynolds analogy is not appropriate for LBE, hence the theoretical analysis and experimental measurements are studied to interpret the flow and heat transfer of LBE in this article. In this way, a single-rod experiment on flow and heat transfer for LBE is performed and a new empirical correlation is fitted. To analyze the behavior of flow and heat transfer for low Prandtl number fluids, and also to validate the turbulence model and new correlation in the hexagonal rod bundles. The CFD simulations are performed for the XJTU 7-rod bundle, the KIT 19-rod bundle, and the ENEA 37-rod bundle based on ANSYS Fluent. The convective heat transfer coefficient of LBE, the temperature of heating wall at different sub-channels, and the temperature non-uniformity are investigated in this paper, to reveal the flow and heat transfer characteristics of LBE. The results show that the deviations of the section averaged Nusselt number are in 25%, and the deviations of the local temperature of LBE and heating wall are in 10℃. Grid spacers will result in a 30% rise in the section averaged Nusselt number, which will improve the turbulent intensity and heat transfer. Temperature non-uniformity is reduced as the Peclet number is increased. The validations demonstrate that the new empirical correlation provides accurate convective heat transfer coefficient predictions for LBE flow and heat transfer in hexagonal rod bundles with a relative derivation in 20%, which can be used as a reference for additional LBE simulations and experiments. [ABSTRACT FROM AUTHOR]

• Heat transfer experiment of hydrogen with variable property was conducted in a heated tube. • New friction and heat transfer correlations were developed with a satisfactory accuracy. • Acceleration parameter was used to modify the heat transfer correlation in a wide range. Hydrogen as a fuel or a coolant plays an essential role in carbon emission reduction and energy system transformation. There has been a long need for a better method of explaining the heat transfer mechanism for gases with variable physical properties in the past few decades. In this study, an experimental investigation was conducted to explore the heat transfer and flow friction of hydrogen in a horizontally heated tube for the following ranges: local Reynolds number from 4300 to 48600, local wall temperature up to 1200 K, local wall-to-bulk temperature ratio up to 2.37, and heat flux up to 0.57 MW/m2. The average friction coefficient equations with heat addition are developed for 6000 < Re < 105 and 1.20 < T w / T b < 1.85. It is found that the modified average heat transfer correlation by the acceleration parameter K v is proposed for 4900 < Re < 44500 and 5 × 10−8 < K v < 1 × 10−6 that can cover the laminar-turbulent transition and turbulence regime with an accuracy of ±10%. A new correlation evaluated at the bulk temperatures is proposed with the temperature and length correction to predict the local heat transfer of hydrogen, which is in good agreement with the experimental data in the range of 8000 < Re < 48000, 6.2 < x / D < 111, and 1.17 < T w / T b < 2.37. This investigation can facilitate improvements in thermal hydraulics analysis codes and the heat transport system with variable fluid properties. [ABSTRACT FROM AUTHOR]

Highlights • A subchannel code based on a homogeneous model was developed and verified. • The thermal-hydraulic characteristics of annular fuel were analyzed. • A burnup-dependent thermal conductivity model was incorporated into the code. Abstract A Subchannel Analysis Code for Annular Fuel (SACAF) capable of modeling dual-cooled annular fuel pins was developed. The coolant flow distribution was adjusted until a uniform pressure drops was formed in all the subchannels. The heat transfer fraction was determined by the criterion that the temperature of the fuel pins was continuous and a burnup-dependent fuel thermal conductivity model was incorporated into the code to analyze the effect of the thermal conductivity degradation. In this paper, the characteristics and verification of the SACAF are described. The thermal-hydraulic performance of the annular fuel was determined for the major thermal-hydraulic parameters and was compared with that of the cylindrical fuel assembly. For the same power distribution, the annular fuel exhibited a greater departure from nucleate boiling ratio (DNBR) and a much lower fuel pellet temperature than the cylindrical fuel. The fuel pin temperature profiles at different burnup levels were obtained and the effects of the thermal conductivity at different burnup levels on the temperature profile and the heat split were determined. [ABSTRACT FROM AUTHOR]

Abstract IVR-ERVC (In-Vessel Retention – External Reactor Vessel Cooling) is a key severe accident management for keeping the integrity of the reactor pressure vessel (RPV) lower head. One of most important topics is the characteristics of the CHF (Critical Heat Flux) on the outside wall of RPV lower head. Some experiments had been conducted to investigate the effect of the orientation on characteristics of the boiling heat transfer and CHF for downward facing heating plate. While the size of the heating plate used may have an appreciable effect on the characteristics of the CHF. In this work, experiments were carried out to investigate the effects of the heater surface area and length-width ratios on saturated pool boiling and CHF for downward facing heating surface. Experimental results indicated that the inclined angle has little effect on the saturated pool boiling heat transfer, and the wall superheat of triggering of CHF was greater for the larger inclined angle. For the same aspect ratio of the heating surface, the CHF decreased with the increase of the heating surface area for the same surface orientation. It also indicated that the CHF was influenced by the bubble escaping length and wetted perimeter complicatedly, and the short edge effect had an appreciable influence on CHF based on present experimental work. Meanwhile, a modified El-Genk and Guo CHF correlation was proposed based on CHF data with the same aspect ratio of the heating surface, which could consider the effect of heating surface area on pool boiling CHF. Highlights • Impacts of downward facing heating surface on CHF. • For same aspect ratio, CHF decreased with the increase of the heating surface area. • CHF was influenced by the bubble escaping length and wetted perimeter complicatedly. • A new CHF correlation proposed could predict experimental data well. [ABSTRACT FROM AUTHOR]

Abstract Nuclear power and thermionic conversion can serve as a compact, durable energy source for the space exploration and exploitation. In this paper, the modified Reactor Excursion and Leak Analysis Program5 (RELAP5) with the implement of NaK-78 eutectic alloy (78%K and 22%Na) properties and heat transfer correlations is adopted to analyze the thermal-hydraulic characteristics of the space nuclear reactor TOPAZ-II. A RELAP5 model including the thermionic fuel elements (TFEs), reactor core, radiator, coolant loop and volume accumulator is established. The temperature reactivity feedback effects of the fuel, TFE emitter, TFE collector, reflector, moderator and the reactivity insertion effects of control drums and safety drums are considered. The steady state condition and three severe transient accidents including reactivity insertion accident (RIA), loss of flow accident (LOFA) and loss of coolant accident (LOCA), are simulated and analyzed. The steady state calculated results agree well with the design values. During the three accidents, the moderator plays a dominant role in the positive temperature reactivity feedback. The coolant has at least 50 K temperature margin to the boiling point. The fuel and TFE components are all below their melting temperature. The progress of these accidents provide relatively sufficient time for operator's response. The calculation results prove that the reactor is a safe and reliable system. Highlights • RELAP5 is modified to analyze a space nuclear reactor TOPAZ-II. • Thermal-hydraulic model of TOPAZ-II is developed and benchmarked. • The steady state and transient accidents of TOPAZ-II are performed. [ABSTRACT FROM AUTHOR]

Abstract Pressurized water reactors (PWRs) use Steam Generators (SG) to transfer the reactor core heat to the secondary loop. The SG contains thousands of heat transfer tubes and tubes plugging operation is always adopted to solve the heat transfer tubes rupture problems caused by the complex operation conditions, including thermal stress, radiation environment and material corrosion. However, tubes plugging operation has great impact on the SG hydraulic performance which leading to some safety issues in nuclear power plants (NPP). In this paper, the AP1000 SG primary side three-dimensional thermal hydraulic simulation model is built using porous media method, in which the important parameters are achieved employing the separate tube simulations. An innovative grid mark method is proposed to realize the flexible tubes plugging conditions. The tubes plugging fractions varies from 5% to 20%, and the effects of tubes plugging locations are also considered. The detailed flow fields in whole AP1000 SG primary side are achieved. The large eddies and some small vortexes are generated in the inlet side of bottom channel head. The momentum loss mechanism is revealed in the whole SG primary loop. Results show that the pressure drop in primary loop increases with the plugging fraction and varies with different tubes plugging positions. The pressure drop goes up to 492.77 KPa in case that the tubes plugging fraction reaches 20%. The tubes plugging position affects the results significantly in case that the plugging fraction is larger than 15%. This work is meaningful for the depth understanding of AP1000 SG tube plugging operations and provides a guideline for the SG maintenance strategy during the whole plant lifetime. Highlights • A three dimensional tube plugging CFD simulation model is built for AP1000 Steam Generator (SG). • An innovative grid marking method is proposed to achieve the different tubes plugging conditions in primary side of SG. • A pressure drop band under different plugging positions is obtained for AP1000 SG. [ABSTRACT FROM AUTHOR]

Highlights • Fluoride salt experimental system has been built using heat pipes. • Mutual interference between heat pipes impedes the heat transfer. • Heat pipe spacing has little difference on the natural convection heat transfer. Abstract In order to improve the passive safety performance of molten salt reactor, a new conceptual design of passive residual heat removal system (PRHRS) has been proposed using heat pipe technology. In this paper, an experimental system has been constructed to verify the PRHRS design of molten salt reactors. An eutectic salt mixture of 46.5 mol%LiF-11.5 mol%NaF-42 mol% KF compositions called FLiNaK is selected as working fluid. Tests on the natural convection between the fluoride salt and a vertical array of heat pipes in the drain tank have been conducted. The temperature field of fluoride salt in the tank with different heat pipe spacing is analyzed. Comparisons have been made on the heat transfer characteristics between the natural convection with a single heat pipe and the vertical array of heat pipes. In addition, the empirical correlation for the Nusselt number is obtained within the range of Rayleigh numbers from 5.7 × 10 6 to 1.31 × 10 7. A good agreement between the proposed correlation and the experimental data is obtained with the deviation of −2.9%–1.7%. This work could provide valuable experimental data and suggestions for the design and analysis of PRHRS of MSR. [ABSTRACT FROM AUTHOR]

• GeN-Foam solver has been extended to water boiling simulation. • OECD/NRC PSBT benchmark is used to preliminarily validate the new solver. • Model sensitivity is analyzed and an empirical turbulent model is developed. • Most of the predictions fall into the ± 2σ exp error bar of the measured data. The OpenFOAM-based GeN-Foam solver has recently been extended to allow simulating water boiling based on a porous-medium approach. This paper describes such extension and preliminarily validates it based on the OECD/NRC PSBT benchmark. Three benchmark exercises are included: steady-state single sub-channel exercise; steady-state rod bundle exercise; and transient rod bundle exercise. Model sensitivity analysis is conducted, with a simple empirical turbulent model that is developed and calibrated to account for the mixing effect caused by the interaction of the two phases. The calculated results are compared with the experimental data, and most of the predictions fall within the ± 2 σ exp error bar. The results indicate that the solver can simulate water boiling heat transfer process and predict void fraction distribution accurately, at least for the cases of the PSBT benchmark. [ABSTRACT FROM AUTHOR]

LBE HCOTSG, which combines the benefits of lead‑bismuth eutectic (LBE) and helical coil once-through tube steam generator (HCOTSG), has the potential to be an excellent heat exchanger choice. Given the similarity of the structure, the analysis of cross/oblique bundle flow heat transfer can help reveal the heat transfer characteristics of the HCOTSG shell side. In this paper, a numerical method for LBE in cross/oblique bundle flow is proposed. Based on the validated method, the application range of the correlation on the flow of the LBE cross/oblique bundle obtained from experiments is broadened based on the numerical simulation work. The influence of tube geometric arrangement in the bundle is also taken into account, including the inclination angle (θ), S min / d, and various staggered arrangements. The difference between numerical simulation results and experimental data is less than 15%. For LBE cross/oblique bundle flow, the extended correlation taking geometric parameters into account also has good applicability. The alternating inclination angle arrangement at an angle of 8° can greatly enhance heat transmission efficacy by up to 33%. The impact of geometric arrangement is estimated in relation to local temperature. This work provides a correlation for the flow of the LBE cross/oblique bundle that is more broadly applicable. Furthermore, this work will significantly benefit the design optimization and program development of LBE HCOTSG or other innovative heat exchangers with cross flow of LBE. • A numerical method for LBE cross/oblique bundle flow is proposed and validated. • The application range of the correlation is broadened based on the method. • Heat transfer performance of LBE flow cross tube bundle is numerically studied. [ABSTRACT FROM AUTHOR]

Among the numerous space nuclear power sources, high temperature gas-cooled reactor with closed Brayton cycle has the advantages of both high power and high energy conversion efficiency. An Open-grid MEgawatt Gas-cooled spAce nuclear reactor (OMEGA) is investigated. Furthermore, a system Transient Analysis code of scheduled Shutdown and emergency Shutdown (TASS) is developed, including core, turbine-generator-compressor, regenerator, condenser, heat pipe radiator and other modules. TASS was verified by comparing prediction value with designed value, and transient performances of system shutdown were simulated. Over the course of shutdown, emissivity of fuel cladding is closely related to cladding peak temperature. Firstly, there is no overlay coating on fuel cladding whose emissivity equals to 0.18, the peak temperature of fuel cladding during scheduled shutdown and emergency shutdown reaches melting point at 7.8 h and 3.6 h respectively. Secondly, there is overlay coating on fuel cladding and emissivity is increased to 0.80. Radiation heat transfer can be greatly enhanced, and peak temperature of fuel pellet and cladding do not exceed the melting point of relevant material, which reflects the importance of fuel cladding surface processing. This paper provides a valuable analysis support for the design of megawatt-class space gas-cooled reactor system. • Shutdown safety analysis of megawatt-class space gas-cooled reactor system is conducted. • Radiation heat transfer among fuel rod claddings is calculated by radiation heat transfer network model. • Emissivity of claddings is the most remarkable contributors at the shutdown process. [ABSTRACT FROM AUTHOR]

• A new method of utilizing the residual heat in reactor using Heat Pipe - Thermoelectric Generator is proposed. • Based on the numerical simulation results, the establishment of the principle prototype is improved. • The principle prototype of HPTEG assembled in reactor was built, and its thermoelectric characteristics were studied. • The thermoelectric performance of the prototype and the influence of temperature on the generating efficiency are obtained. This paper proposes a waste heat utilization method for new type reactor by Heat Pipe - Thermoelectric Generator (HP-TEG) to utilize the waste heat of lower divertor in fusion device and stirling generator in heat pipe space reactor. The feasibility of the HPTEG waste heat utilization device is verified by its principle prototype. Firstly, the numerical simulations of the single heat pipe and principle prototype were carried out to provide a numerical optimization scheme for the design of the principle prototype. The simulation results show that the heat transfer efficiency of the copper–water heat pipe is about 93%, while the overall heat transfer efficiency of the principle prototype is about 85%. Based on the simulation results, the thermal stress concentration areas were found, mainly concentrated in the thermal insulation short of the heat pipe. During the construction of the principle prototype, stainless steel sleeves were installed outside the insulation section of the heat pipe to limit the thermal expansion caused by the thermal stress concentration. Besides the principle prototype of HPTEG assembled in new type reactor was built, and its thermoelectric characteristics were studied, the results verified the feasibility of using waste heat. The test results show that the principle prototype starts smoothly, and the heat pipe has suitable isothermal properties (△T < 15 ℃). Due to the heat pipe's high heat transfer efficiency, the heat pipe obtained from the experiment is 91%–96%. The principle prototype has good heat transfer efficiency, which is 81.94%–93.25%. The maximum total electric power output of TEG is 16.37 W, in which the corresponding current is 0.52A, and the power generation efficiency is up to 1.49%. This paper verifies the feasibility of HPTEG to utilize the waste heat of lower divertor in fusion device and stirling generator in heat pipe space reactor, which provides a reference for the utilization of reactor residual heat. [ABSTRACT FROM AUTHOR]