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The compact design of the marine nuclear power reactor results in a significant reduction in the size of the main equipment, particularly the steam generator. The primary steam-water separator is the core component in this regard. The swirl vane separator is the primary separator in the steam-water separator, responsible for more than 80% of the steam-water separation task. The performance of the marine nuclear power system directly affects its safety and economy. The primary focus of this study centered on the swirl vane separator, employing the Euler-Euler two-phase flow model in conjunction with the RNG k-ε turbulent flow model. Furthermore, an additional inertia force resulting from the rolling motion was integrated into the momentum equation as a source term using user-defined functions (UDF). The impact of varying gravity components was considered, leading to the development of a three-dimensional numerical computational model to analyze the flow and separation of steam and water within the separator. The impact of varying rolling angle and rolling period on the flow properties and operational efficiency of the separator was methodically investigated. The results indicate that the turbulent mixing of the steam-water two-phase fluid within the separator is enhanced by the rolling motion, leading to a chaotic pressure field, velocity field, and distribution of liquid volume fraction. The critical state is observed when the rolling angle reaches 40° and the rolling period is 2 s. This critical state causes fluid backflow in different parts of the separator, greatly affecting the effectiveness of separating the two-phase flow. The additional inertia force from the rolling motion results in periodic pressure and velocity fluctuations within the separator, with a phase delay of one-quarter period compared to the separator’s motion. In the near-wall region of the separation cylinder, the liquid volume fraction reaches its peak when the separator is at its maximum inclination angle. The periodic changes in separation efficiency follow a sinusoidal pattern. Before reaching the critical condition, the variation in separation efficiency corresponds with the rolling motion of the separator. The influence on the separation performance becomes more pronounced with higher rolling magnitude and shorter rolling period.

The deterministic two-step method for the fast reactor neutronics calculation, composed of cross-section homogenization and diffusion or transport core calculations, was widely applied in the fast reactor engineering design and analysis field. The homogenized cross-section calculation method based on Monte Carlo with continuous-energy and fine geometry can provide high-precision cross-sections for advanced fast reactors. The current development status and trends of coupling Monte Carlo-generated homogenized cross-sections with diffusion and transport core calculations were briefly reviewed in this paper. The methods discussed include the Monte Carlo flux-volume homogenization method, the superhomogenization equivalence technique (SPH), and the Monte Carlo flux-moment homogenization method (MHT). The MET-1000 metal fuel fast reactor is used as a benchmark. The SPH equivalent techniques are widely used to preserve the reaction rates of a reference heterogeneous model and a homogenous model. In this paper, the SPH was applied to the control rods' cross-section address to improve the diffusion core calculations. This equivalence technique reduces the overestimation of the control rod worth using the diffusion core solver from 13.5% to 0.35% and improves power distribution prediction accuracy. With SPH correction, the MC/diffusion in this work exhibits about

The steam generator tube rupture (SGTR) accident is one of the design basis accidents of lead-cooled fast reactor (LFR). In the event of such an accident, a subcooled water jet is generated in the lead pool of primary circuit, which forms a complicated distribution of the liquid metal-gas two-phase flow. The detection of the two-phase interface and migration of steam bubbles is an significant part of reactor safety analysis. Due to the opacity, corrosiveness, high temperature of liquid metal, and the current limitations of measurement methods involving liquid metal-gas two-phase flow, a diamond-like carbon (DLC) coated wire-mesh sensor (WMS) based on electrical principles was proposed in this paper. The sensor could achieve the image of the liquid metal-gas two-phase distribution inside the flow channel, while preventing the corrosion failure of the sensor. Based on the COMSOL software, a numerical model of the DLC coated WMS was established. Selecting the materials of two phases as liquid GaInSn alloy and air, the simulation of the electric field in the liquid metal environment containing a single bubble was performed, and an electric signal value matrix representing the liquid metal-gas two-phase was output. The signal matrix was further normalized through the linear relation to obtain the two-phase distribution image, which verified the application feasibility of the coated WMS in liquid metal. In addition, a prototype of the DLC coated WMS was manufactured, and the numerical simulation results were experimentally validated in order to demonstrate the rationality of the numerical model. Furthermore, the influence of different structural parameters on the measurement accuracy of coated WMS was studied in detail by changing the transverse interval, axial interval, electrode diameter, coat thickness, and the size of air bubble. The simulation results show that the electric field distribution in the liquid metal-gas two-phase environment differs significantly from that in conventional fluid medias such as water-gas due to the high relative permittivity of liquid metal. On the other hand, considering the resolution of the WMS and the phenomenon of electric signal crosstalk, the transverse interval of electrode should be selected between 2-3 mm, and the axial interval should be selected between 1.5-2 mm to achieve high measurement accuracy. The coat thickness and electrode diameter have little influence on the measurement accuracy, but to reduce the intrusion effect of the WMS, the electrode diameter should be selected between 0.4-0.5 mm and the coat thickness should not exceed half of the electrode radius. This study could guide the practical application of DLC coated WMS in the liquid metal-gas two-phase environment and provide a technical basis for the analysis of multi-phase distribution under SGTR accidents.

In the development of critical heat flux (CHF) correlation, the mixing factor β (i.e., the turbulent thermal diffusivity) is a characteristic parameter representing the cross-mixing effect between sub-channels. At present, the mixing factor obtained through the cross-mixing experiment is essentially the macroscopic equivalent quantity of averaging the cross-mixing effect (such as turbulent cross-mixing, flow sweeping, etc.). In fact, the mixing effect caused by the grid will decrease along the downstream, which will lead to a certain deviation in the local parameters calculated by the average mixing factor. The influence of the deviation on the DNBR limit value of the CHF correlation and the safety margin is not clear. Thus, the CHF correlation matching with the sub-channel analysis program FLICA-ⅢF was developed by using the minimum DNBR point method. Owen criterion was used to determine the DNBR limit. After achieving the development methods of CHF correlation, the prediction effect, DNBR limits and actual maximum power of correlations of different mixing factors were compared and analyzed. The results indicate that in the range of experimental conditions, the local pressure of minimum DNBR point does not change with the change of mixing factor, the local mass flow rate increases with mixing factor, and the local quality decreases with the increase of mixing factor. The deviation of mixing factor will increase the DNBR limit of its correlation, but there is no linear relationship between the two. Besides, in actual conditions, the maximum power calculated by the correlations of different mixing factors is different. While the maximum power value of the precise mixing factor is also the largest. In conclusion, the more precise mixing factor determined by experiment has higher accuracy and economy in fitting correlation. Therefore, it is valuable and meaningful to consider the fine subchannel of grid mixing effect.

The helically metallic fuel is promising to improve the power density and safety margin of reactors core by its advantages of high thermal conductivity, large heat transfer area-to-volume ratio and continual inter-channel mixing. However, helical geometry and metallic U-Zr fuel will introduce challenges in the analysis of neutron physics, thermohydraulics and mechanics performance. Thus, in the current work, the analysis methods were developed for the characteristics of helically metallic fuel in neutron physics, thermohydraulics, mechanics and multi-physics coupling. For neutron physics, a 3D continuous-energy Monte Carlo method was employed to address the complex geometries and the complex intermediate neutron energy spectrum and generate few-group cross-sections. In core calculations, the 3D method of characteristics (MOC) with high geometric flexibility, accuracy and convergence was utilized. This method has been verified in 3D calculations of helically metallic fuel core. The results demonstrate that both cross-section generation and core calculation have achieved high accuracy and effectively improved computational efficiency. For thermohydraulics, the experimental technique including visual measurement and wire mesh sensor were used to measure the mixing characteristics of single and two-phase flow in helically metallic fuel rod bundle. CFD and subchannel analysis codes were developed to predict the boiling and critical heat flux. The inter-channel mixing was dominated by the flow sweeping mixing, while the turbulent mixing was insignificant. The vapor phase crowded at the elbow of the rods, where the boiling crisis was triggered. For mechanics, the molecular dynamics method was employed to predict the fundamental thermal-mechanical properties. The correlations were proposed for the thermal conductivity and elasticity modulus of U-Zr alloy under irradiation conditions with pores. The finite element tool Abaqus was used to analysis the thermal-mechanical performance of helically metallic fuel rod bundle. The maximum stress was found at the blade tips where adjacent rods contact with each other. The maximum stress varied from 326.7 MPa under fresh condition to about 313.3 MPa at 14.1% FEMA because of creep, which was far lower than the failure strength of the rod. The structure integrity can be ensured during the full life cycle. Based on the individual analysis method, the multi-physics coupling framework combining the neutron physics, thermohydraulics and mechanics was developed for the design and performance evaluation of the fuel assembly and reactor core with helically metallic fuel rods. Based on this multi-physics analysis tool, a novel boron-free small modular pressurized water reactor NETH-HCF175M design using helical-cruciform metal fuel was proposed. The core can achieve a cycle length of about 1 360 EFPD as the reactivity of the core is completely controlled by burnable absorbers and control rods.

Corrosion of structural materials in lead-bismuth fast reactors due to liquid lead-bismuth eutectic (LBE) is a significant issue that hinders their development. To slow down the corrosion rate of high temperature LBE on structural materials, it is crucial to precisely control the oxygen concentration in LBE, which can form a stable protective oxide layer on the surface of the structural materials. The high density of LBE means that its erosive effect on structural materials during flow cannot be ignored. To investigate the dynamic corrosion characteristics of T91 steel cladding tubes in a high-temperature LBE environment, a corrosion model that couples oxidation, reduction, and erosion was established. The computational fluid dynamics (CFD) method, along with the SST k-ω turbulence model and lead-bismuth turbulence Prandtl number model were used. The simulation analysed the thermal-hydraulic characteristics of the fuel assembly, as well as the corrosion phenomenon of the fuel cladding surface. The analysis identifies the key factors that affect corrosion. The results indicate that at a specific flow velocity of LBE, the equilibrium thickness of the spinel layer increases with the rise in surface temperature of the cladding along the liquid LBE flow direction in the fuel assembly. Additionally, the thickness loss rate of the cladding also increases. At the end of 100 hours of service, the upper cladding of the fuel rod is covered by a double oxide layer consisting of magnetite and spinel. At the end of 800 hours of service, only the magnetite layer remains at the exit of the fuel assembly rod cladding. The surface of other cladding has only one spinel layer. As the liquid LBE flow velocity at the fuel assembly inlet increases, the cladding temperature decreases, but the cladding thickness loss rate increases due to the increase in wall shear stress. When the inlet flow velocity is 2 m/s and the oxide layer is stable, the center rod’s cladding thickness loss rate is 0.044 98 mm/a. When the oxygen concentration on the surface of the cladding exceeds the minimum value required for the oxidation reaction, the cladding thickness loss rate increases with rising surface temperature. Conversely, when the oxygen concentration on the surface of the cladding is below the minimum value required for the oxidation reaction, the cladding dissolves in liquid LBE at a rate that can reach thousands of mm/a. This study proposes a numerical simulation method for predicting LBE core oxidation corrosion characteristics. The method can be used to accurately predict the corrosion of LBE cores.

BackgroundNatural circulation systems are widely used in the nuclear industry for decay heat removal, post-accident containment cooling, and the cooling of radioactive waste storage facilities. Because a large number of spent fuel assemblies are densely arranged in the spent fuel pool, the spent fuel assemblies in the central area of the spent fuel pool can be approximated as having adiabatic boundary conditions. In the event of a coolant loss accident, the heat generated by the spent fuel assemblies can be exported only by the natural circulation of air.PurposeThis study aims to develop analytical models for fuel assembly natural circulation characteristics independent of the results of computational fluid dynamics (CFD) to investigate the natural circulation characteristics of spent fuel pools following a coolant loss accident.MethodsBased on the experimental results of a fuel assembly pressure drop of a full-scale pressurized water reactor (PWR), the Darcy–Forchheimer model was firstly revised to predict the pressure drop of humid air flowing through the fuel assembly. Models for the fuel assembly natural circulation flow rate and peak cladding temperature were then established, and the effects of the relative humidity of air on the models were considered. Finally, these models were applied to investigating effects of total heating power, ambient temperature, and relative humidity on the natural circulation flow rate and peak cladding temperature of the fuel assembly.ResultsThe results show that the models accurately predict the natural circulation flow rate and peak cladding temperature of the fuel assembly under different heating powers, and the errors are less than 25% and 20%, respectively, as compared with the experimental measurement values.ConclusionsThe developed models in this study can be used to study the natural circulation characteristics of full-scale PWR fuel assemblies.

BackgroundSteam generator tube rupture (SGTR) accidents in lead-bismuth cooled fast reactors result in the generation of numerous steam bubbles owing to the interaction of the high-temperature liquid lead-bismuth eutectic (LBE) from the primary circuit and high-pressure subcooled water of the secondary circuit. These steam bubbles carried by the LBE may enter the core, causing local heat transfer deterioration and a power transient, seriously affecting reactor safety.PurposeThis study aims to elucidate the movement and dynamic behaviors of steam bubbles in liquid LBE and develop a drag coefficient model applicable to bubble migration in LBE for assessing the safety of the core during SGTR accidents.MethodsBased on the coupled level-set and volume-of-fluid (CLSVOF) method, a three-dimensional numerical model for calculating the drag coefficient of steam bubbles in LBE was established to study their movement and dynamic behaviors. Firstly, the deformation, velocity, and trajectory characteristics of bubbles in LBE were analyzed, and their drag coefficients were estimated. Then, the simulated drag coefficient values were compared with those from existing models whilst the prediction performance of the model proposed by Tomiyama was superior to those of the others. Finally, Tomiyama's drag model was optimized by introducing the We number, and its applicability was analyzed.ResultsAnalysis results show that the errors of the existing drag coefficient models are large under the condition of bubble–LBE two-phase flow, and the calculation error of the optimized model for the bubble drag coefficient in LBE is within 15%.ConclusionsThe feasibility of drag calculation model optimized in this study is verified, demonstrating its suitability for calculating the drag coefficient of steam bubbles in LBE.

Reactor fuel assemblies may be biased and bowed as a whole due to factors such as manufacturing and installation deviations or external forces. This consequence may have an impact on heat transfer and criticality, and even the possibility of edge-rod criticality. In order to evaluate the impact size, a set of high-quality dynamic mesh generation technology was developed through CFD. The dynamic mesh technology was used for the base condition to make the overall displacement or bowing of the fuel rods. Using the Eulerian two-fluid model and the improved RPI wall boiling model, the CHF of the 5×5 rod bundle channel with different offset distances and bowing degrees was predicted, and the CFD prediction results were verified with the CHF look-up table value. The results show that the CHF predicted by the non-deformation and small deformation conditions are in good agreement with the values of the sub-channel parameter CHF look-up table, while the large deformation condition triggers the side rods critical and CHF look-up table is not applicable. Because the CHF table is made based on an 8 mm round tube, the non-8 mm round tube or rod bundle structure (such as the central sub-channel) can be corrected by the hydraulic diameter correction factor, and the prediction accuracy is often high. However, at the edge and corner channel position of the rod bundle assembly, the heated boundary and wet circumference are not uniform, and the offset or bowing of the fuel rods squeezes the corner channel, resulting in a more unbalanced distribution of the heated boundary and wet circumference at this location. As a result, the CHF look-up table is not applicable. The critical occurrence position and numerical value under different deformation conditions were compared, the distribution of mass fluxes in different sub-channels at the critical height was also compared, and the sensitivity of various deformation types to CHF was summarized in this paper. For the calculation results of the undeformed fuel assembly, the flow rate of the central sub-channel is the highest, the side channel is the second, and the corner channel is the smallest. For the deformed fuel assembly, the offset and bowing change the flow distribution of the sub-channels, the flow of the corner channels decreases, and the flow of the central sub-channel increases. When the gap of the corner channel is so small that the coolant flow cannot effectively take away the bubbles on the surface of the fuel rods, causing local congestion of the bubbles, the side rods will be critical. The XY-type condition has the strongest influence on the corner channel flow, but the weakest effect on the overall flow distribution, and the C-type bowing condition has the strongest influence on the overall flow distribution. In the calculation condition, the axial power is uniformly biased and bowed by 0.5 mm without changing the critical axial position, while bowing by 0.8 mm and 1 mm affects the critical position. The larger the three deformation degrees, the smaller the predicted CHF value and the increase of the critical probability of the side rods. Among them, the XY-type deformation is the least sensitive to the CHF value, while the C-type bowing with a cosine axial heat flux has the greatest impact on the CHF value. Compared with the same degree of bowing, the critical heat flux density value of axial power cosine is smaller. When bowing 0.5 mm, the critical position of the axial power uniform condition occurs at the center, while the critical position of the axial power cosine condition occurs at the edge. That is, the cosine power may also be the cause of the criticality of the side rods.

As a new type of reactor fuel assembly, helical-cruciform fuel (HCF) assembly has some advantages. The cruciform section shortens heat conduction path and the twist structure strengthens inter-channel mixing, which can upgrade core power density with enough safety margin. The periodic self-supporting structure fixes the assembly instead of spacer grid to reduce flow resistance. However, because of the small contact area at self-supporting structure the stress concentration is prone in contact position, which has a negative effect on the integrity of cladding. To solve this problem, the thermal-fluid-force coupling characteristics of typical HCF assembly were studied in this research. The flow and heat transfer of coolant were simulated by ANSYS Fluent to obtain velocity, temperature and vapor fraction distributions at single-phase and two-phase flow. Steady-State Thermal and Static Structural module in ANSYS were used to analyze the thermal and mechanical behavior of HCF assembly, respectively, and the differences of stress and strain distributions between single-phase flow and two-phase flow were found. The thermal-fluid coupling was two-way, and the data were exchanged in System Coupling module. Because the deformation of HCF assembly was too small to change flow and heat transfer of coolant, the thermal-force coupling was one-way. The results show that the cross flow intensity of coolant near the cladding is greater than that in the center of channel, and the direction of vortices at the two places is opposite. The cross flow affects the coolant temperature and vapor fraction distribution and makes the vapor fraction fluctuate with a period of 45° along flow direction at two-phase flow. The temperature at outer surface of cladding and pellet center also fluctuates with a period of 90° along flow direction except for the entrance part at single-phase flow and two-phase flow. Because a small amount of bubbles at the cladding surface can enhance the heat transfer, the heat transfer coefficient at two-phase flow is twice as large as that at single-phase flow. The maximum Mises stress of 300 MPa is in the lobe region of HCF assembly with 0.1 plastic strain. The thermal stress of HCF assembly in the lobe region is mainly related to the contact constraint, but in the valley region it is mainly related to temperature gradient inside HCF. At the cladding of HCF assembly, the Mises stress in the valley region and the plastic strain in the lobe region at single-phase flow and two-phase flow are the same, but in the lobe region the Mises stress at two-phase flow is 30 MPa smaller than that at single-phase flow.

BackgroundPassive residual heat-removal system (PRHRS) based on natural circulation has been widely used in small reactors connected on secondary side loop.PurposeThis study aims to develop and analyze a passive residual heat-removal system on the secondary side of a small reactor to improve reactor safety.MethodsBased on the completed heat-transfer experiment of the intermediate circuit of a small integrated nuclear power plant, Reactor Excursion and Leak Analysis Program (RELAP) was employed to determine and analyze the natural circulation characteristics of the intermediate loop were determined and analyzed.ResultsThe heat transfer rate results of the program are in good agreement with the experimental data, and the natural cycle characteristics of PRHRS can be characterized. The pressure of the system loop is determined by the average temperature of the primary side of the steam generator (SG), and the difference in the inlet temperature of the SG primary side, mass flow rate, and height of the cooling and heating sources significantly influence the heat-transfer performance of the SG system. The heat-transfer performance of the waste-heat-removal system is more sensitive to the resistance of the system loop when the temperature of the primary-side inlet of the SG is higher.ConclusionsThese results provide a valuable application for further investigations of passive systems in small reactors.

The helical-coiled once-through steam generator (H-OTSG) is widely used in small modular reactors due to its compactness and higher heat transfer efficiency. In this study, an experimental and numerical simulation study of the thermalhydraulics characteristics of HOTSG under different thermal power conditions was carried out, and the experimental results were used to verify the onedimensional system code SGTH1D. The configuration consists of 85 helically coiled tubes divided into ten layers according to the coil diameter. Both steadystate experiments and flow instability experiments were conducted under thermal power condition ranging from 06 MW to 23 MW. Firstly, the steadystate experimental results show the difference between the shellside inlet temperature and the tubeside outlet temperature increases with thermal power, which is due to the decreasing of the heat transfer capacity of the HOTSG. Sensitivity analysis of the system parameters shows that the shellside pressure has little effect on the average heat transfer coefficient. And the average heat transfer coefficient also is insensitive to tubeside pressure and shellside inlet temperature under low thermal power conditions. However, under low thermal power conditions, the average heat transfer coefficient increases with the decrease of the tubeside pressure and the increases of the shellside inlet temperature. Secondly, the flow instability experimental results show that the increase of the thermal power can stabilize the HOTSG and decrease the inlet throttling of the flow instability thresholds because the length of the subcooled singlephase region increases with the thermal power. The sensitivity analysis of the system parameters shows that the increase of the tubeside pressure and the decrease of the shellside inlet temperature are beneficial to the stability of the system. The influence of the shellside parameters on the flow instability threshold is weaker than that of the tubeside parameters. Finally, a numerical study on the thermal-hydraulics characteristics of the configuration was carried out in the present study. By comparing the experimental data with the numerical simulation results of the SGTH1D code, the accuracy of the SGTH1D code in predicting the steadystate heat transfer characteristics and flow instability of configuration was verified. The SGTH1D code can accurately predict the heat transfer rate of the HOTSG under various thermal power conditions, and the prediction errors of shellside and tubeside outlet temperatures are within ±1 ℃. The numerical results of flow instability characteristics of the SGTH1D code are conservative at low thermal power condition, and the error of numerical results is within ±20% when the thermal power is larger than 12 MW.

As the pressure and heat transfer boundaries between primary and secondary sides, the steam generator (SG) should be integrated during operation. However, SG suffers from the problems including vibration, wear, fatigue and corrosion, which will lead to the degradation of tubes and other components and may lead to the failure of SG. The research on the degradation of SG is associated with many disciplines, including thermohydraulics, structural mechanics, water chemistry and materials science, among these, the thermohydraulics are the foundation of other analysis. In the current work, SGTHPorous3D, a 3D steam generator thermohydraulics analysis code based on twofluid porous media model, was developed for the prediction of 3D thermohydraulics in the Utube SG (UTSG) and helical coiled SG (HCSG). First, the twofluid porous media model for the straight tube, U-bend and helical tube bundle was proposed for the general application of UTSG and HCSG. A velocity projection method was employed to estimate the flow resistance and heat transfer for inclined sweeping flow over different types of tubes. With these models, the 3D numerical strategy based on coarse mesh was developed for the shell side of SG. Second, as for the tube side flow, the parallel 3D multichannel model was used to simplify the numerous spatially distributed tubes. The effects of centrifugal force were considered by employing the flow resistance and heat transfer maps specially developed for Ubend and helical shaped tubes. Finally, a data mapping method was also developed for the tubetoshell side heat transfer between cells of different spatial scales. A partial relaxation scheme was used to accelerate the convergence of coupled heat transfer, where the heat flux was relaxed while the temperature wasn’t. The experiment data from the Westinghouse MB2 UTSG, the KAERI HCSG and the SJTUNETH HCSG were used for code validation. The SGTHPorous3D code was applied to predict the 3D thermalhydraulics parameters in the CAP1400 SG. The distributions of thermal-hydraulics parameters for each phase and the mixture phase were obtained. High non-uniformity can be observed for the flow and heat transfer parameters. The ratio between heat released in the hot side and cold side is as large as 2.7. The crossflow energy related to the flowinduced vibration is also derived to find the possible location of tube damage. The most probable location for a potential tube rupture locates at approximate 04 rad in the hot side. The results show that the SGTHPorous3D can predict the 3D thermalhydraulics parameters in both of the tube and shell sides of UTSG and HCSG, which provides support for the heat transfer design and flowinduced vibration analysis for the SGs.

Helically coiled tube (HCT) is widely used in nuclear engineering and other fields due to its effective compactness, excellent structure robustness, heat transfer efficiency and flow stability. As one of the most important gasliquid twophase flow patterns in HCT, the characteristic of annular flow has a great influence on the heat transfer performance of helical oncethrough steam generator. Especially, the circumferential distribution of liquid film thickness in annular flow directly determines the dryout onset location. Although the liquid film flow in straight tubes has been extensively studied, due to the limitation of measurement technology, the precise measurement data of liquid film thickness in HCT are still very rare. In this paper, based on the selfdeveloped noninvasive contact ringisland array sensor (RIAS), airwater twophase flow experiments were carried out to study the threedimensional spatialtemporal distribution of liquid film thickness in vertical HCT with different structures. According to the results, four typical liquid film flow regimes were defined based on the circumferential position of the thicker liquid film appearing in the tube wall. That is, the bottom distribution (BD) dominated by gravity force, the outside distribution (OD) dominated by liquid centrifugal force, the inside distribution (ID) dominated by gas centrifugal force and the insideoutside distribution (IOD) dominated by secondary flow. The effects of coil diameter and coil pitch of HCT as well as gas and liquid velocities on the distribution of liquid film flow regimes were detailed analyzed. It is implied that the coil diameter determines the intensity of centrifugal force and secondary flow. Generally, the smaller the coil diameter, the easier the circumferential distribution forms of OD, ID and IOD are to appear. Also, increasing the coil pitch weakens the effect of centrifugal force on the liquid film. Then a small coil pitch normally results in the circumferential distribution form of OD while it is ID for large coil pitch. It is also found that the IOD occurs only when both gas and liquid velocities are large enough. By taking the influence of HCT geometry and fluid parameters into account, a general submap of liquid film flow in HCT was proposed using the modified Dean number De* and the modified LockhartMartinelli parameter X*. The transition criteria between different liquid film flow regimes were also established and verified with experimental data in present work and previous references. From this work, the mechanism of curvatureinduced centrifugal force and secondary flow on the spatiotemporal evolution of liquid film in HCT is revealed.

Non-uniform cross flow goes across the heat exchanger tubes of steam generators in nuclear plants, resulting in serious vibrations and wears. The pitch velocity varies among tube bundles and each tube is subjected to different fluid forces, causing inconsistent vibration response of each tube. To get the flow induced vibration characteristics of the whole heat exchanger tube bundle, an experimental setup of threespan tube bundle with 28 circular tubes and 10 semicircular tubes was designed. The test section has a glass window at the top for vision measurement. Coded reference points were attached on the top of the heat exchanger tubes to improve the accuracy of vision measurement. Different codes correspond to different tubes, enabling the image processing program to automatically identify each tube. The center dots of the reference points also provide a stable tracking target for the image processing program. A baffle with 1/3 opening was set upstream to create nonuniform cross flow to the tube bundle. Vibration displacement of each tube was measured with a high-speed camera and vibration acceleration of a typical tube was measured by an accelerometer, while the tube bundle was subjected to this nonuniform cross flow. In order to contrast with nonuniform cross flow conditions, the vibration of the tube bundle facing a uniform orifice baffle was tested. The results show that the heat exchanger tube bundle is more prone to fluidelastic instability (FEI) under nonuniform cross flow conditions compared to uniform orifice baffle conditions. And the vibration amplitude and frequency of each tube of the whole bundle is unique under nonuniform cross flow conditions. Four local resonance regions coexist in the tube bundle when flow induced vibration instability happens. The most significant fluidelastic instability resonance happens to the central tubes at the bottom three columns facing the opening of baffle. The peak frequency of vibration in this region is consistent with the firstorder natural frequency of the tube with one fixed support in middle and one simple support on top. The non-uniform distribution of the tube bundle vibration response will lead to inconsistent wear of the tubes located in different regions, which adversely affects the maintenance and service life of steam generators. The FEI critical velocity under the nonuniform cross flow was compared with the results predicted by five empirical correlations. Connors correlation and Chen correlation are conservative. The FEI critical velocity predicted by Connors correlation and Chen correlation are 502% and 247% lower than the experimental value, respectively. To predict the FEI critical velocity more accurately under nonuniform cross flow conditions, the Gorman correlation, which is closest to the experimental value, is modified. The results provide guidances for the design of heat exchanger tube bundles of steam generators.

An experimental study is conducted to find the characteristics of steam plume and pressure oscillation on direct contact condensation by a side-hole sparger. Synchronal measuring of transient pressure of the steam plume is gained from the high-speed camera and high-pressure sensor respectively. The influence of steam mass flux and water temperature on direct contact condensation characteristic are presented and its regime map is plotted. Then, the dynamic connections of transient pressure and steam plume in different condensation regimes are analyzed. It is found that high frequency pressure oscillation and the collapse of detached bubbles occur at the same time. Together with the condensing and disappearing process of the collapse of detached bubbles, the intensity of pressure oscillation decays exponentially in a vibration way. The changing trends of steam plume length in condensation oscillation regime and stable condensation regime are also obtained, which shows that the steam plume increases with the steam mass flux and the temperature in the condensation oscillation regime. When entering the stable condensation regime, the steam plume suddenly decreases and then increases with the temperature and the steam mass flux. The research results are useful for the engineering application of sparger in steam emission devices.

LBE-cooled fast reactor (LBFR) has a promising application prospect in energy generation and transmutation of highlevel radioactive waste. Wirewrapped fuel rods are usually used in LBFR to reach high heat transfer efficiency and selfsupport. The wire spacer has a great influence on the transverse flow and interchannel mixing, and leads to a nonuniform temperature profile over the fuel rod surface. Therefore, accurate prediction of the transverse flow mixing characteristics in the wirewrapped bundle channel is the basis of thermalhydraulic safety analysis of LBFR. In the present study, a CFD modeling and simulation for a LBEcooled 7pin wirewrapped fuel assembly was carried out based on RANS method. The simulation results were firstly compared with the experimental data from the literature. A good agreement demonstrated the capability of the employed approach to predict the flow field in the rod bundle. Then, the threedimensional transverse flow mixing behaviors of LBE in single wirewrapped fuel and multi wire-wrapped fuel assemblies were analyzed in detail. The results show that the maximum transverse flow velocity in the central sub-channel of the single wirewrapped fuel assembly does not exceed 19% of the axial flow velocity. The transverse flow direction and secondary flow center change periodically with height. In the single wirewrapped fuel assembly, the transverse flow at the interface of the central subchannel is distributed as a trigonometric function along the axis. When the wire spacer coincides with the interface of the central subchannel，the transverse flow and the transverse flow mixing index tend to zero. When the wire spacer is perpendicular to the interface, the transverse flow and the transverse flow mixing index reach the peak. The transverse flow mixing index is not sensitive to Reynolds number in the turbulent zone, but is strongly correlated to the structural parameters of the assembly, which means that the lateral mixing in the wirewrapped bundle is dominated by the flow sweeping mechanism, rather than turbulent mixing. Compared with the single wire-wrapped fuel assembly, the transverse flow distribution of the multi wirewrapped fuel assembly is more uniform and more secondary flow vortex can be observed. The swirling directions of the LBE around the two adjacent fuel rods in the multi wirewrapped fuel assembly are opposite. That is, there are two opposite transverse flow directions on the interface of the central subchannel in the multi wirewrapped fuel assembly.

A full-scale novel conical swirl-vane steam separator with a conical barrel is studied by test and numerical methods using steam and water as working fluids. The test results indicate that the separation efficiency of the separator decreases first and then increases with the increase of steam superficial velocity. But this phenomenon is not obvious at low steam superficial velocities. The pressure drop of the separator increases significantly with the increase of steam superficial velocity. Based on the commercial software STAR-CCM+ and Euler-Euler two-fluid model, a numerical simulation is conducted to study the effect of droplet size on the separation efficiency and pressure drop of the swirl-vane separator. The numerical results imply that large droplets promote the liquid separation. However, the pressure drop is insensitive to the droplet size. According to the test results, a numerical model that is in good agreement with the test is established, and the distribution of presssure, velocity, and water volume fraction are analyzed in detail.

Due to the compact structure, high heat transfer efficiency and strong reliability, helical coil once-through steam generator (HCSG) is widely used in advanced small reactors. In HCSG, the liquid in the tube side is heated by the helical coils so that subcooled water in the tube side is transferred into superheated steam, and the entire heat transfer regime is covered. Moreover, the evolution mechanism of thermalhydraulic behaviors is complex because it is affected by the helical geometry, the centrifugal force and the secondary flow. Hence, the development of advanced HCSG was limited due to the lack of more accurate models of heat transfer and regime transition. With the financial support of the three major nuclear power groups in China, experimental and theoretical research on heat and mass transfer behaviors in helical coils was carried out systematically. Different helical coils were investigated and comprehensive parametric analysis of heat transfer was conducted to provide a better understanding of different heat transfer regimes. According to our research, a fundamental database of heat transfer and flow resistance in helical coils with a wide range of geometric dimensions and thermalhydraulic parameters was established. In addition, the transition pattern of the circumferential nonuniform heat transfer mechanism was clarified. Due to the secondary flow, it consequently leaded to the nonuniform distribution of flow field. The heat transfer and flow resistance characteristics were notably affected by the special behaviors of liquid in the helical coils. Hence, based on mechanism analysis, several models were proposed to construct a high precision library of allregime prediction model, such as a threeregion dryout quality model, a postdryout heat transfer model based on dryness gradient and a widerange model for twophase frictional multiplier factor. For the model of dryout quality, it takes the relative intensity of gravity, buoyancy and centrifugal force into consideration. Since different dominant forces lead to different heat transfer behaviors, a piecewise correlation is established to give more accurate predictions. As for the postdryout heat transfer model, it considers the influence of dryness gradient and achieves a more accurate prediction. For the twophase pressure drop model, the modified frictional multiplier factor is adjusted according to our database. Then, a prediction model was proposed which was also validated by other researchers and had great prediction accuracy in a wide range. In all, this work includes part of our investigation on the heat transfer of helical coils and provides a highprecision prediction in the entire heat transfer regime, which meets the design requirement of HCSG for advanced reactors.

BackgroundAs an innovative nuclear fuel assembly, the helical cruciform fuel (HCF) assembly has the characteristics of large specific heat transfer area, short heat conduction path, strong inter-channel mixing and free from the grid spacers. Compared with the traditional cylindrical fuel assembly, the HCF assembly can raise the core power density with compromise on the safety margin. However, the concentrated stress might take place at the location of self-support points, resulting in the plastic deformation and even rupture.PurposeThis study aims to analyze the thermal-mechanical behaviors of HCF bundle under steady conditions and accident transitions, so as to obtain the stress and strain of HCF rods, based on which, the integrity of fuel cladding was assessed.MethodsFirstly, a 3×3 typical HCF geometrical assembly model without four rods in corners was constructed and discretized by hexahedral mesh. Then, the steady and transient convective conditions were applied to the outer surfaces of rods to simulate the various working conditions, including single phase, boiling, reactivity insertion accident and loss of coolant accident. Finally, the governing equations for mechanics and heat transfer were established and solved in ANSYS using the thermal and mechanical modules.ResultsThe results show that, the maximum von Mises stress and plastic deformation take place at the location where adjacent rods contact, where the stress and strain are determined by both the contact constrain condition and the temperature difference between cladding inner and outer surfaces. However, at the elbow of the blades, the stress and strain are mainly affected by the radial temperature gradient in the cladding material. For the cladding, the plastic deformation is larger while the von Mises stress is smaller under the flow boiling condition compared with these under the single-phase cooling condition. Furthermore, the integrity of fuel cladding can be maintained under the conditions of reactivity insertion and loss of coolant accidents, where the stress and the temperature are lower than the break limit and the zirconium-water reaction temperature, respectively.ConclusionsFrom the thermal-mechanical analysis on the HCF assembly, this kind of innovative fuel assembly shows good mechanical performance under normal and accidental conditions.

BackgroundWhen steam generator tube rupture (SGTR) occurs in the lead-bismuth cooled fast reactor, high pressure water/steam flows into the primary side filled with high temperature liquid metal. According to the location and size of the rupture, the leakage behaviors of the rupture may involve leak-before-break (LBB), single-phase critical flow or two-phase critical flow. Under the action of high temperature liquid metal, different forms of heat and mass transfer behaviors occur in the two-component multiphase system of water-metal, which has an important influence on the safe operation of the lead-bismuth cooled fast reactor.PurposeThis study aims at the bubble dynamic behavior in the descending flow field of liquid lead-bismuth alloy (LBE) in the tube bundle in different stages of SGTR caused by microcrack on the surface of the heat transfer tube during the drying stage and low flow single phase steam permeates the primary side.MethodsBased on the VOF method, a numerical simulation model of steam-LBE two-phase flow and phase interface capture was established to study the bubble growth and transport behaviors from single tube or 3×3 tube bundle in the downward flow field of high temperature LBE. The SST k-ω model was employed to solve the turbulence equation. The physical law of steam bubble movement was analyzed and its influence on the heat transfer and operation stability of steam generator was evaluated.ResultsThe results show that the dynamics behaviors of steam bubbles in the descending flow field are quite different from these in static liquid or upward flow. The steam bubbles may slide along the heat transfer tube surface after departure from the crack under the actions of LBE descending flow field and buoyancy. The steam bubbles may form a steam film covering the heat transfer tube surface or accumulate by quantity in the bundle.ConclusionsThese phenomena adversely affects the flow stability of the LBE and the heat transfer of the steam generator.

BackgroundAs a nuclear power system with high energy density and strong adaptability, heat pipe cooled reactor can play an important role in space and deep sea exploration, and is attracting more and more attention from all over the world. Due to the solid-state core and high-temperature operation of the heat pipe cooled reactor, the thermal expansion and thermal stress are non-negligible. The nuclear-thermal-mechanical coupling effect should be considered for the heat pipe cooled reactor core.PurposeThis study aims to investigate the thermal-mechanical characteristics of the reactor core, and evaluate the safety performance of the reactor under typical conditions and transient accidents for a typical heat pipe cooled nuclear reactor, Nuclear Silence ThermoElectric Reactor (NUSTER).MethodsBased on FLUENT and MECHANICAL modules in finite element analysis software ANSYS, calculation of the nuclear-thermal-mechanical coupling was conducted for the reactor core of NUSTER under the steady-state and the reactor reactivity insertion accident respectively, hence the temperature and stress distribution were obtained. The core design safety was evaluated by comparing the temperature and stress peak values of each component with the melting point and yield strength of material.Results & ConclusionThe comparison results show that NUSTER core can operate safely within the safety limit in both steady state and accident conditions. Each component has a safety margin of 200 K for temperature and 150 MPa for stress under accident condition, so the core design safety has been verified.

Objective: To compare the predictive value of Oxford acute severity of illness score (OASIS) and simplified acute physiology score II (SAPS II) for in-hospital mortality in intensive care unit (ICU) patients with sepsis., Methods: A retrospective cohort study was conducted using the data in the Medical Information Mart for Intensive Care-IV 0.4 (MIMIC-IV 0.4). Based on Sepsis-3 diagnostic criteria, the basic information of ICU adult sepsis patients with infection and sequential organ failure assessment (SOFA) score ≥ 2 within 24 hours of ICU admission admitted for the first time in the database was extracted, including gender, age, vasopressor drugs, sedative drugs, mechanical ventilation, renal replacement therapy, length of ICU stay, OASIS, SAPS II scores, etc. The primary outcome was in-hospital mortality. A receiver operator characteristic curve (ROC curve) was drawn, and the area under the ROC curve (AUC) was calculated to compare the prognostic value of OASIS score and SAPS II score., Results: A total of 11 098 adult ICU sepsis patients were enrolled in the final analysis, of which 2 320 died and 8 778 survived in hospital, with a mortality of 20.90%. Compared with the survivors, the non-survivors were older [years old: 71 (60, 81) vs. 67 (56, 78)], had longer length of ICU stay [days: 6.95 (3.39, 13.07) vs. 4.23 (2.19, 9.73)] and higher proportions of using vasopressor drugs, sedative drugs, mechanical ventilation and renal replacement therapy [vasopressor drugs: 50.65% (1 175/2 320) vs. 33.05% (2 901/8 778), sedative drugs: 58.53% (1 358/2 320) vs. 48.41% (4 249/8 778), mechanical ventilation: 89.57% (2 078/2 320) vs. 81.66% (7 168/8 778), renal replacement therapy: 11.98% (278/2 320) vs. 6.57% (577/8 778), all P < 0.01]. Moreover, the non-survivors had higher OASIS score [43 (36, 49) vs. 35 (29, 41), P < 0.01] and SAPS II score [49 (40, 60) vs. 38 (31, 47), P < 0.01] as compared with the survivors. ROC curve analysis showed that the AUC of OASIS score and SAPS II score for predicting in-hospital death of ICU patients with sepsis was 0.713 [95% confidence interval (95%CI) was 0.701-0.725] and 0.716 (95%CI was 0.704-0.728), respectively, and the Delong test showed no significant difference in AUC between the two scoring systems (P > 0.05)., Conclusions: OASIS score has a good predictive value for in-hospital mortality in sepsis patients, which is similar to SAPS II score. OASIS score is simpler and has a broader clinical application prospect than SAPS II score.