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AbstractInteraction of prevailing ocean waves and wind with the platforms containing the small modular reactors (SMRs) employed in marine environments may significantly alter the flow and friction characteristics inside these reactors. The present research is focused on the numerical study of the effects of rolling motions on the turbulent flow and frictional characteristics of a three-dimensional closed loop of narrow rectangular channels using Ansys Fluent. The computational results have been corroborated with experimental data present in literature. The results illustrate that flow and friction characteristics fluctuate sinusoidally as the loop undergoes rolling motion. Strong fluctuations were observed in the flow rate and time-dependent friction coefficient with an increase in rolling amplitude or reduction in the rolling period. These variations became more pronounced at low Reynolds numbers and diminished at higher Reynolds numbers. Increasing the fluid viscosity also subsided the rolling effects. The average flow velocity in the loop was found to decrease from 0.27 to 0.15 m/s in various phases of the rolling period. The relative Reynolds number was found to be reduced by 50% under rolling motions for the range of steady-state Reynolds numbers investigated in the present study. The transient friction coefficient was also found to oscillate under rolling motion with the same period as that of excitation. The transient friction coefficient’s oscillations also increased with rolling amplitude or reduction in the rolling period. However, the temporally averaged friction coefficient under rolling motions was found to be equal to the steady-state frictional coefficient in the loop. [ABSTRACT FROM AUTHOR]

• Concept of "LESS is MORE" is proposed in the design of lead-bismuth cooled reactor. • Core design of lead-bismuth cooled small modular reactor is conducted. • Neutronic and thermal-hydraulic performances of reactor core are analyzed and improved. • Preliminary core design of LBE cooled reactor could satisfy the design criteria and is reasonable. Small modular reactors (SMR) are in great demands in the near future and the lead-based fast reactor is one of the most important directions in the development of SMR. In this paper, core design and analysis of a LEad-bismuth Small MOdular Reactor (LESMOR) was developed and examined to realize the notion of "LESS is MORE". LESMOR is a 300 MWt pool-type reactor where all facilities are mounted in a reactor vessel. Advanced nitride fuel is adopted, and the spent-plutonium is used as the driven fuel and thorium is used as the fertile fuel. Sub-channel analysis was performed in the assembly design using an in-house sub-channel code SUBAS, and the 11 × 11 scheme with a pitch-to-diameter (P/D) ratio of 1.4 was adopted. The full core neutronic performance was studied and evaluated in this paper. Results show that the reactor can be maintained critical for 20 years and the maximum radial power factor is 1.28. The steady-state thermal-hydraulic performance of the core was analyzed, including the characteristics of the mass flow distribution and the hottest assembly. The peaking fuel cladding temperature is 519.2 °C, and the maximum fuel center temperature is 723.4 °C, both within the temperature limit. This paper provides valuable information in the neutronic and thermal-hydraulic design and analysis of LBE nuclear reactor. [ABSTRACT FROM AUTHOR]

An enhanced moving particle semi-implicit method for simulation of incompressible fluid flow and FSI (Fluid-Structure Interaction) is developed in the present paper. Based on the recent studies on LSMPS (Least Square Moving Particle Semi-implicit) method [1,2] , the fluid models for FSI are improved, and the elastic structure models and fluid–structure coupling schemes are developed. Many strategies are developed to reduce the influence of mechanical vibration of elastomers on fluid calculation. The models are validated step by step. The improved fluid models with different Dirichlet and Neumann boundary are validated through hydrostatic pool problem, connecting tube problem and dam break problem. The elastic structure models under the velocity boundary and the loading boundary are validated through free oscillation cantilever problem and pressurized ring problem respectively. The fluid-structure coupling schemes are validated by the violent sloshing problem without considering the elastic deformation and the problem of dam break with an elastic gate. The results show a good performance of the developed method on the simulation of incompressible fluid flow and fluid-structure interaction. [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]

The lead-based fast reactor (LFR) is one of the most promising choices of the Generation IV reactors as it has better transmutation and breeding capability, as well as higher safety and economical efficiency. In this paper, the new concept LFR Pb-Bi cooled direct-contact-boiling water fast reactor (PBWFR) is studied. Coupled thermal-hydraulic analysis with the in-house system analysis code SACOL and sub-channel analysis code SUBAS is conducted using the one-direction multi-step coupling method. The unprotected transient of over power (UTOP) and unprotected loss of heat sink (ULOHS) accidents are simulated, and the key parameters are compared with the results by single SACOL, preliminary demonstrating the accuracy and rationality of the coupling method. The transient analysis shows the major concern for PBWFR’s operation is the cladding temperature. In UTOP, the maximum peak cladding temperature (PCT) reaches 834.15 °C, exceeding the cladding temperature limit of 800 °C, which suggests the reactor would be at a risk during UTOP. The maximum PCT is 679.43 °C in ULOHS. Due to the immediate reactor shutdown in ULOHS, the safety margin is relatively larger but the PCT is still higher than the limit in normal condition. Therefore, it is better to further optimize the structure or the operation condition of PBWFR. In addition, more detailed information of the core thermal hydraulic performance can be obtained. In this way, the coupled code could make up the shortage of single channel models and extend the function of the sub-channel analysis code. [ABSTRACT FROM AUTHOR]

Summary: Pb‐Bi‐cooled direct contact boiling water fast reactor (PBWFR) featured with a direct‐contact heat exchanger between lead‐bismuth eutectic coolant and water could significantly simplify the primary system and enhance the natural circulation capability, meeting the potential needs for small modular reactor design. It is of great importance to conduct thermal‐hydraulic analysis of the PBWFR core in detail. In this paper, a self‐developed SUB‐channel AnalysiS code SUBAS is adopted to study the thermal hydraulic characteristics of the PBWFR core. The fidelity and the reliability of the code have been preliminarily benchmarked. With SUBAS, the space grid is studied to figure out its impact on the temperature and flow distributions in each sub‐channel. Besides, the application of space grids would increase the pressure drops and decreases the cross flow between adjacent sub‐channels. To study the transient performance of the PBWFR core, the power transient and the inlet blockage accident are calculated by SUBAS. The results of the power transient show the cross‐flow effect would be weakened in the sub‐channel which has higher coolant temperature and larger mass flow rate. For the inlet blockage accident, the results indicate the influence of the small area blockage is relatively weak on the overall performance of the assembly but is significant on the local parameters. With consideration of time and space, the blockage influence only exists in a certain area. This research may provide contribution to the design of PBWFR. [ABSTRACT FROM AUTHOR]

In this study, a three-step approach of ordinal 3D-1D&2D-3D is independently proposed to derive a coupling optimization design rule of the Chinese Fusion Engineering Test Reactor (CFETR) helium cooled solid breeder (HCSB) blanket. Firstly, the three dimensional (3D) neutronics model of CFETR is built with all of the 15 blanket modules adopting the original radial build arrangement parameters. Based on this model, both the neutron wall loading (NWL), which will serve as the normalization factors for the following local neutronics calculations, and the tritium breeding ratio (TBR) distributions on each blanket module are obtained. Secondly, the radial build arrangements of all the 7 inboard and the 8 outboard blanket modules in an 11.25° toroidal CFETR sector are optimized by taking both the tritium breeding and thermal performances into account as the optimization objectives. NTCOC, a N eutronics/ T hermal-hydraulic C oupling O ptimization C ode, which has been independently developed by Xi’an Jiaotong University for accelerating the radial build design and optimization of the fusion reactor HCSB blanket concept with adopting the simplified 1D neutronics and 2D thermal-hydraulic models, is used for these work. Thirdly, based on the final optimized schemes of all the blanket modules, a comprehensive research on the coupling optimization design rule of the CFETR HCSB blanket is performed in-depth and a coupling optimization design integration approach has been finally summarized. Finally, the 3D CFETR neutronics model is updated by replacing all of the original blanket radial build arrangement parameters with the final optimized schemes. The calculation results show that after the radial optimizations, the total TBR of all the 15 blanket modules has increased from the previous 1.092–1.274, which can well meet the self-sufficiency tritium breeding requirement of CFETR. [ABSTRACT FROM AUTHOR]

Reliable systematic uncertainty analysis is of great significance for reactor safety and licensing. In this paper, efforts have been made to quantify uncertainties of computational results. The uncertainty quantification and sensitivity analysis for a 20 MWth prismatic Transportable Fluoride-salt-cooled High temperature Reactor (TFHR) proposed by MIT is performed with the DAKOTA code coupled to the best-estimate nuclear system software RELAP5-3D. Eight input parameters are selected with specified variations and probability density functions (PDFs), a total number of 400 steady-state and Unprotected Loss of Flow (ULOF) transient calculations are then conducted in order to statistically obtain the 5% and 95% peaking fuel temperature tolerance intervals with confidence level of 95%. The Latin hypercube sampling (LHS) model in DAKOTA is verified against analytical results and the TFHR RELAP5 modeling is benchmarked based on CFD calculations. It is found that, for the steady-state, the uncertainties of the heat transfer coefficient and helium gap conductivity factor are the most remarkable contributors to the statistical results of peaking fuel temperature. 0.5% of calculated fuel temperatures exceeded the steady-state fuel temperature limit of 1573 K. Under ULOF accident condition, no temperature points exceeded the transient fuel failure limit 1873 K, but there is a possibility of 0.195 leading to the damage of structural material. Three primary contributors to the uncertainty in peak fuel and outlet coolant temperature are the reactor power, helium gap, and heat transfer coefficient. The findings indicate that there are needs for further reactor design optimization from thermal-hydraulics prospective. [ABSTRACT FROM AUTHOR]

The application of thorium fuel in advanced CANDU (CANadian Deuterium Uranium) reactors is an economic solution to sustainable development of nuclear energy. In this paper, the thermal hydraulic design and transient analysis of Thorium-based Advanced CANDU Reactor (TACR) heat transport system are conducted using self-programmed code CANTHAC (CANDU Thermal Hydraulic Analysis Code). In the code CANTHAC, the single channel model and the point kinetics model are adopted to simulate the core thermal hydraulic and neutronic behaviors. A distributed parameter model is adopted in the simulation of integral economizer U-tube steam generator (IEUTSG). Multi-region non-equilibrium model is applied to simulate the pressurizer. The characteristics of heat transport pumps were simulated by four-quadrant analogy curves. The code is verified with ACR-700 steady-state analysis. The calculation results agreed well with ACR-700 design values and the relative errors are all within 2%. With CANTHAC steady-state module, the preliminary thermal hydraulic design of TACR is conducted. A 1000 MW double loop reactor with 520 fuel channels is designed. Single loop mass flow rate was designed to be 3202 kg/s. Under the design scheme, the maximum cladding temperature and the maximum temperature are 340.2 °C and 1369.9 °C, which satisfied the design requirements. The transient analysis of TACR heat transport system was performed with CANTHAC transient module. SG feed water temperature reduction accident, and complete loss of Class IV accident were conducted. In the SG feed water temperature reduction accident, a new balance of neutron and thermal hydraulic is reached and the core power increases by 5% at last. In the complete loss of Class IV accident, the maximum fuel temperature reaches peak value at 1172 °C and still has some safety margin. The ROH (Reactor Outlet Header) pressure does not exceed 105% of the design value, which meet the safety criterion demand. [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 characteristics of LBE under rolling conditions are investigated experimentally. • The time-averaged flow frictional resistance of LBE is increased under rolling conditions. • The transverse additional force is main reason for the increase of flow resistance. • The results may be useful for the safety analysis under moving conditions. The effects of rolling motion on the flow characteristics of lead-bismuth eutectic in a vertical circular tube are investigated experimentally. The results show that the pressure drop and flow rate of the forced LBE circulation fluctuate under rolling conditions. The effect of rolling motion on the flow characteristics decreases with the increase of the corresponding static Reynolds number. The rolling motion may increase the time-averaged frictional resistance coefficient of the LBE when flow rate is low. The critical value of the impact Reynolds number is about 125,000. As a result, the time-averaged flow rate decreases. The analysis shows that it is mainly the transverse additional force caused by the rolling motion increases the frictional resistance of the LBE. An empirical correlation is established to predict the frictional resistance coefficient under rolling conditions. This study may be useful in the design and safety analysis of LBE as a working fluid under moving conditions. [ABSTRACT FROM AUTHOR]

Free-surface flow coupled with solid-liquid phase change can be observed in many industrial applications, and it has a great influence on many industrial processes. In this paper, the Incompressible Smoothed Particle Hydrodynamics (ISPH) method is extended with solid-liquid phase change model. A novel implicit viscosity solver is adopted in which the implicit viscosity solver is moved behind the calculation of pressure force to eliminate the time step limitation and the numerical creeping of high viscosity. To reduce the compression caused by the delayed incompressibility correction, accurate wall boundary conditions are applied and an enhanced implicit viscosity Laplacian operator is proposed. Besides, a dynamic solid boundary criterion is proposed to deal with the changing solid-liquid interface. In addition, the enthalpy-based heat transfer model and the mushy zone viscosity model are introduced. The inter-particle conductivity between liquid and solid is adopted to consider the large contact thermal resistance between fluid and solid. As will be shown, our ISPH method is verified in its performance in simulating free-surface flow coupled with solid-liquid phase change by four numerical examples and has good advantages in simulation quality over the existing implicit viscosity algorithm, which, hopefully, is to provide applications in engineering. • ISPH method is developed for free-surface flow coupled with solid-liquid phase change. • Enhanced implicit viscosity algorithm is proposed for high viscosity flow. • Dynamic solid boundary criterion is proposed for the changing solid-liquid interface. • The contact thermal resistance model and the mushy zone viscosity model is introduced. • Numerical examples verify the method and demonstrate the improvements. [ABSTRACT FROM AUTHOR]

As one of three candidate tritium breeding blanket concepts for Chinese Fusion Engineering Test Reactor (CFETR), a conceptual structure of helium cooled solid breeder (HCSB) blanket was recently proposed. In this paper, the preliminary thermal-hydraulic and safety analyses of the typical outboard equatorial blanket module (No.12) have been carried out using RELAP5/Mod3.4 code. Two design basis accidents are investigated based on the steady-state initialization, including Loss of Off-Site Power and In-Box Loss of Coolant Accident (LOCA). The differences between circulator coast down and circulator rotor locked under Loss of Off-Site Power are compared. Regarding the In-Box LOCA, the influences of different break sizes and locations are thoroughly analyzed based on a relatively accurate modeling method of the heat structures in sub-modules. The analysis results show that the blanket and the combined helium cooling system (HCS) are designed with sufficient decay heat removal capability for both accidents, which can preliminarily verify the feasibility of the conceptual design. The research work can also provide an important reference for parameter optimization of the blanket and its HCS in the next stage. [ABSTRACT FROM AUTHOR]

• The CFD-DEM method is a new method developed for the formation of debris bed from irregularly shaped debris. • The simulation for DEFOR-E7 is performed and analyzed with the debris size obtained by the SEINA code. • The deposition morphology of a debris bed with a height of 0.1 m, a repose angle of 35° and a porosity of 72 % in DEFOR-E7 test are obtained. In a severe accident in nuclear reactor, the core meltdown occurs when nuclear fuel rods cannot be sufficiently cooled, thus forming debris beds with specific shapes on the lower head of the Reactor Pressure Vessel (RPV). In this paper, the fluid–solid interaction during debris bed formation is studied based on Computational Fluid Dynamics method coupled with Discrete Element Method (CFD-DEM). The CFD models are used to simulate the flow and pressure field of fluid, the DEM models are used to calculate the collision between solids and the collision between solids and wall, as well as the fluid–solid interaction force can be calculated by the two-way coupling models. The DEM models and CFD-DEM models were validated by different experiments. Based on the above models, the DEFOR-E7 test was simulated, and a conical debris bed formed by irregular debris with a static angle of 35° and a porosity of 72% was obtained. Therefore, the method and findings in this paper are significative on theory and application for the study of nuclear severe accidents. [ABSTRACT FROM AUTHOR]

A system analysis code coupled with the heat pipe model is developed to analyze the thermal-hydraulic characteristics of the improved TOPAZ-II reactor power system with a heat pipe radiator. The core thermal-hydraulic model, neutron physics model, and the coolant loop component models (including pump, volume accumulator, pipes and plenums) are established. The designed heat pipe radiator, which replaces the original pumped loop radiator, is also modeled, including two-dimensional heat pipe analysis model, fin model and coolant transport duct model. The system analysis code and the heat pipe model is coupled in the transport duct model. Steady state condition and start-up procedure of the improved TOPAZ-II system are calculated. The results show that the designed radiator can satisfy the waste heat rejection requirement of the improved power system. Meanwhile, the code can be used to obtained the thermal characteristics of the system transients such as the start-up process. [ABSTRACT FROM AUTHOR]

Summary: To meet the requirements of the unmanned underwater vehicles (UUVs) for the power source, a conceptual design of a 100‐kWe level Nuclear Silent Thermal‐Electrical Reactor (NUSTER‐100) is proposed. The NUSTER‐100 is a 1‐MWt reactor that has been designed to couple with a 100‐kWe thermoelectric generator system. The high temperature sodium heat pipes are adopted to cool the reactor core. The cascaded wide temperature range thermoelectric generation (TEG) system is used to achieve thermoelectric conversion, which ensures high power generation efficiency and low operating noise. The neutronic and thermal‐mechanical analysis results are provided for the design. The core power distribution and energy spectrum are obtained. The neutronic analysis results are imported into ANSYS as input for thermal‐mechanical analysis. In this paper, the fuel‐HPs channel and the 1/8 symmetrical core are selected as the analysis objects, and the temperature distribution and statics parameters under the normal operation condition are obtained. The analysis results show that under normal operation conditions, the core parameters meet the design requirements. These design parameters are all lower than the safety limit. This work provides a certain reference for the design and application of the heat pipe reactor power system. [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]

Thermal stratification caused the significant temperature differentials from top to bottom inside the pipe that imposed fluctuating high stresses on the wall of the pipe and eventually could lead to severe structural integrity concerns and degrade fatigue strength. This study aimed to evaluate the transient thermal distribution and fatigue damage caused by thermal stratification in the pressurizer surge line (SL) of Nuclear Power Plants (NPP). In this paper, the effects of nonuniform temperature distributions caused by thermal stratification were investigated by performing a detailed Computational Fluid Dynamic (CFD) study using the Large Eddy Simulation (LES) approach. The fluid flow was simulated using a full buoyancy model and conjugate heat transfer at the pipe inner wall and the fluid interface. Numerical results of the investigation in the SL pipe were validated with similar experimental setup data obtained from a prototype experimental facility of a pressurized water reactor. Fatigue damage due to induced thermal stresses in the SL model was investigated and fatigue hotspots were also identified. This work provides a comprehensive case study of thermal stratification for estimating pressurizer surge pipeline structural integrity. [ABSTRACT FROM AUTHOR]

Summary: Heat pipes utilize the latent heat of vaporization and lie on the capillary forces to maintain the circulation of working fluid. High‐temperature heat pipes applied in the nuclear silent thermal‐electrical reactor are investigated and optimized. Nine wick structures and several design variables are investigated in the single/multi‐objective optimization. The improved network thermodynamic model of a heat pipe is adopted, and the thermal resistance, mass, and transport capacity of wicks are selected as the objective functions. Multi‐objective optimization results are obtained by Pareto‐optimal points. Comparative results reveal that mesh screen and annular artery present a better performance (0.00180 K/W) in both single and multiple objective optimizations. Potassium is preferred for the wick of the circular artery and rectangular artery, while sodium is preferred for the wick of isosceles‐triangular groove and sintered fibers. And with the operating temperature (800 ~ 950 K), thermal resistance is decreased with the simultaneous rise of mass (165.2%) and transport capacity (154.6%) in the mesh screen. This work makes it possible to improve the operational performance of a heat pipe and provides a reference for the selection of wick structure and heat pipe design. [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]

Advanced small modular reactors (SMRs) use different design in the systems, structures, components from large reactors for achieving a high level of safety and reliability. In present work, the SMRs severe accident caused by the station blackout (SBO) was modeled and analyzed using MELCOR code, and the simulation of the accident scenario response to SBO was conducted. Based on the steady state calculation, which agrees well with designed values, we introduced the SBO accident for transient calculation. First, the case of the SBO accident without the passive core cooling system (PXS) was calculated. The progression and scenario in the reactor pressure vessel (RPV) and the containment were simulated and analyzed, including the transient response, cooling capacity and thermal-hydraulic characteristics and so on. The station black-out transient in the SMR can be simulated accurately, and the main failure model in the accident process can be concluded. Then three other cases of the SBO accident with different passive safety systems (core makeup tank (CMT), accumulator (ACC), passive residual heat removal system heat exchanger (PRHR HX), automatic depressurization system (ADS)) of the PXS were calculated respectively, and the results for different passive safety systems were compared. The passive core cooling system can not only provide water to the primary coolant system, but also take away the reactor decay residual heat. So in a station black-out transient, we can get more time for restoring AC power, and effectively prevent the accidents such as Fukushima. [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]

In recent years, Computational Fluid Dynamics (CFD) methodology has been widely used for solving multi-dimensional thermal-hydraulics problems in nuclear science and engineering. As for the neutronics analysis, it mostly depends on the deterministic or probabilistic method codes which are independent of the thermal-hydraulic ones. Usually an extra interface program needs to be developed for the data exchange between the two kinds of codes in the coupled calculation. In this paper, a Temporal And Spatial Neutronics Analysis Module (TASNAM) in commercial CFD package FLUENT was developed based on the User Defined Function (UDF) and User Define Scalar (UDS) to solve the neutronics problems. Neutron diffusion equations and delayed neutron precursors balance equations are discretized by finite volume method (FVM) and can be solved with CFD conservation equations simultaneously. In order to demonstrate the feasibility and accuracy of TASNAM in FLUENT, three neutronics benchmark problems, the 2D-IAEA, 2D-TWIGL, and 2D-LRA benchmark problem are analyzed. The calculation results in the present paper have a good agreement with the reference values, which demonstrates that the TASNAM in FLUENT can solve neutronics problems accurately. [ABSTRACT FROM AUTHOR]

There are many differences between the flow and heat transfer characteristics of nuclear reactors under ocean and land-based conditions for the effects of ocean waves. In this paper, thermal hydraulic characteristics of a passive residual heat removal system (PRHRS) for an integrated pressurized water reactor (IPWR) in ocean environment were investigated theoretically. A series of reasonable theoretical models for a PRHRS in an IPWR were established. These models mainly include the core, once-through steam generator, nitrogen pressurizer, main coolant pump, flow and heat transfer and ocean motion models. The flow and heat transfer models are suitable for the core with plate-type fuel element and the once-through steam generator with annular channel, respectively. A transient analysis code in FORTRAN 90 format has been developed to analyze the thermal–hydraulic characteristics of the PRHRS under ocean conditions. The code was implemented to analyze the effects of different ocean motions on the transient thermal-hydraulic characteristics of PRHRS. It is found that the oscillating amplitudes and periods of the system parameters are determined by those of the ocean motions. The effect of rolling motion is more obvious than that of pitching motion when the amplitudes and periods of rolling and pitching motions are the same. The obtained analysis results are significant to the improvement design of the PRHRS and the safety operation of the IPWR. [ABSTRACT FROM AUTHOR]

The isothermal steam oxidation experiments of Cr-coated Zr alloy cladding at 1350 ℃ were conducted, and the influence of Cr-Zr eutectic reaction on the oxidation behavior was investigated. The oxidation duration, ranging from 50 s to 3600 s, covers the cases from the initial formation of eutectic liquid to the complete oxidation of the cladding. The microstructures of the coating and the substrate were characterized. The eutectic-oxidation degradation mechanism of the coating is summarized. The migration of the eutectic liquid and the evolution of Cr-rich phases in the post-oxidation eutectic structure are studied and discussed, confirming the process of ZrCr 2 reacting with α-Zr(O) to form Zr 3 Cr 3 O. The degraded Cr coating and the Cr retained in ZrO 2 during the migration of the eutectic liquid can hardly affect the oxidation of cladding substrate. However, the Cr-rich phases, which eventually aggregated in the radial center of the cladding, can reduce the oxidation rate. [Display omitted] • Systematic isothermal steam oxidation experiments of Cr-coated Zr alloy cladding at 1350 ℃ were conducted. • Microstructural evolution mechanism of Cr coating and Zr substrate induced by eutectic and oxidation reactions is summarized. • The formation of Zr 3 Cr 3 O during the oxidation of Cr-Zr eutectic liquid is confirmed. • The oxidation kinetics are summarized and the influence of Cr-rich phases is 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]

• This study utilizes the increasing flow regime experimental data to extend the prediction range and improve the accuracy. • Integrated flow regime prediction models considering vertical and horizontal flow direction were developed and evaluated. • The MLP, RF, and SVM algorithms were utilized to develop the flow regime prediction model respectively. • The flow regime prediction model shows better performance than existing models in three directions respectively. The nuclear reactor core is the pivotal component in a nuclear power plant to generate and transfer heat, so accurate prediction of reactor core flow and heat transfer characteristics is a crucial problem for the reactor system design. Flow regime is closely related to the thermal–hydraulic characteristics of two-phase fluid. Still, flow regime models used in thermal–hydraulic calculation codes rely heavily on earlier experimental data, featuring a narrow application range and unsatisfactory accuracy. To extend the prediction range and improve the accuracy of flow regime prediction by making full use of the increasing experimental data, integrated flow regime prediction models considering vertical flow direction for light water reactor (LWR) and horizontal flow direction for Canadian Natural Deuterium Uranium (CANDU) were developed and evaluated based on three standard machine learning algorithms in this study. Firstly, the experimental data of horizontal and vertical flow collected from the literature was modified and preprocessed to be training data. Then, the multi-layer perceptron (MLP) algorithm, random forest (RF) algorithm, and support vector machine (SVM) algorithm were utilized to develop the flow regime prediction model respectively. The performance of the three models was evaluated and compared and the results showed that the flow regime prediction model based on RF was the optimal model with higher prediction and more efficiency than the other two models. Finally, the flow regime prediction model was compared with existing models in three directions respectively, which indicated that the range and accuracy of the model based on RF were superior to the existing models. This study provides an implantable and scalable approach for flow regime prediction, and the application range and accuracy of flow regime prediction can be continuously improved with updating training data. [ABSTRACT FROM AUTHOR]

• The transient response of FuSTAR in five design basis accidents are analyzed with the transient simulation code FUSY. • The sensitivity analysis is performed for some key parameters that may affect the transient behavior. • FuSTAR has a great ability to survive in URIA and ULOF. • The maximum limit of external parameters causing ULOHS and UOC are obtained by sensitivity analysis. FuSTAR is a small, modular fluoride-salt-cooled high-temperature advanced reactor, proposed by Xi'an Jiaotong University (XJTU), aiming at fulfilling the requirements of Gen IV nuclear system and meanwhile ensuring the inherent safety and economics. In order to evaluate the safety performance of FuSTAR, five typical design basis accidents are selected for safety analysis with different initial events, which are unprotected reactivity insertion accident (URIA), unprotected loss of flow (ULOF), unprotected loss of heat sink (ULOHS), unprotected overcooling (UOC), and the accident combing ULOF with ULOHS. The sensitivity analysis is also performed. FuSTAR is modeled by modified system analysis code and the simulations of the steady-state are verified by the results with CFD method. The results of transient accidents indicate that safety of FuSTAR has been guaranteed in URIA and ULOF. Meanwhile, the maximum limit of external parameters causing ULOHS and UOC are obtained by sensitivity analysis. In the ULOF and ULOHS, the peak of fuel temperature, cladding surface temperature and core outlet temperature of coolant are below their safety limits. Therefore, FuSTAR exhibits acceptable safety under design basis accidents. This study will provide supports in verifying the rationality of the design and the subsequent optimization of FuSTAR. [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]

Fuel pins assembled in Pressurized Water Reactor (PWR) and Liquid Metal-cooled Fast Reactor (LMFR) core are usually cooled by single-phase coolant such as water or liquid metal. Researches on the coolant flow and heat transfer phenomena through rod bundles is quite essential for the nuclear reactor core design. This paper briefly summaries the current developments of Computational Fluid Dynamics (CFD) studies on the thermal hydraulic analysis of single-phase flow in the rod bundles of nuclear reactor core. On account of the tremendous amount of publications related to each aspect above, this brief review would focus on the most representative researches in the past decades. The aim of this paper is to provide an outline and present a comprehensive overview of the CFD research and guidelines on single-phase flow through bare rod bundle rod with mixing vane spacer and wrapped wire spacer in reactor core. In addition, the analysis and validation methods for various CFD calculating methods are concluded. The applications of CFD approach with different type of rod bundles are summarized. Finally, some remaining challenges for future work are proposed and discussed. [ABSTRACT FROM AUTHOR]

In this study, we investigate the steam oxidation behaviors of molten zirconium alloy at exceptionally high temperatures of 1870 ℃ and 1970 ℃. A specialized high-temperature steam oxidation apparatus was established to conduct the steam oxidation experiment. The experiments yielded crucial data including oxidation weight gain results, macroscopic appearance of the oxide surface, and section morphology. Several new phenomena: α-Zr(O) precipitates and two-layer oxide structure, were observed in this study. Based on these observations, a complete mechanism of molten zirconium oxidation in steam was developed, providing an interpretive framework for the oxidation observed in this experiment. This newly proposed mechanism reevaluates the formation process of nonceramic precipitates in the molten bulk and its consequential impact on oxidation kinetics at temperatures below 2000 ℃. • High-temperature steam oxidation behaviors of molten zirconium alloy at 1870℃ and 1970℃ were studied. • The oxidation mechanism of molten zirconium was discussed. • The enhanced effect of the melt on the oxidation kinetics was analyzed. [ABSTRACT FROM AUTHOR]

• Two coupled parallel frameworks for shell and tube heat exchangers are proposed. • Fully three-dimensional parallel computation on both sides of the heat exchanger is realized. • One-to-one method achieves 85% parallel efficiency in 14 cores computation. • One-to-many method improves computational efficiency by about 20 times over serial in 28 cores computation. Shell and tube heat exchanger is widely used as a common type of heat exchanger. When the size and the number of heat exchanger tubes are enormous, it is difficult to use the direct modeling method for numerical simulation of heat exchangers. The simplified modeling method based on porous media utilizing user-defined function (UDF) has been widely applied. However, when the three-dimensional (3D) modeling is required on the tube side, the parallelization problem still restricts the computational efficiency of numerical simulation. In this paper, a one-to-one parallel method was first developed to solve the problem of coupled parameter transfer between the primary and secondary sides in the parallel framework, which was tested to achieve a parallel efficiency of 85% under 14 cores. However, the method acquires complicated data transfer between nodes, large memory occupation, and the modeling of the tube side is different from the actual structure under certain circumstances. To solve the problems, a one-to-many parallel method is proposed, and the computational efficiency is tested to be enhanced by about 20 times under 28 cores compared with single core, with a parallel efficiency of 70%. The two methods proposed in this paper are of great significance to the parallel transformation of the future 3D full-size code for shell and tube heat exchangers development and the calculation efficiency improvement. [ABSTRACT FROM AUTHOR]

• A Hi-fi space thermionic nuclear reactor system model is established, and its dynamic performance is obtained. • A model predictive controller is designed for the space reactor system using discrete state-space model and Kalman filter. • Several transient scenarios were designed to evaluate the control performance of the model predictive controller applied to space reactors from various perspectives. • Stable and rapid control of the space reactor has been achieved. Space nuclear reactors possess the inherent benefits of independence from solar energy and the capability to withstand intricate external circumstances. Distinguishing themselves from terrestrial reactors, space nuclear reactors function within the realm of outer space, operating autonomously throughout their entire operational cycles. Consequently, the establishment of a dependable automatic control system assumes utmost significance. In this paper, a high-fidelity nonlinear space thermionic nuclear reactor model is established, encompassing neutron physics, thermal–hydraulic, and thermoelectric conversion system models. An automatic control system with a model predictive controller (MPC) is designed based on the system. A simplified state-space model is established based on system identification as the internal model for the MPC controller, demonstrating good agreement with the dynamic response of the nonlinear system model. The deployment of the MPC controller for system control simulations exhibits excellent control performance under various operating conditions. In the case of a nuclear power step response, the MPC controller achieves an overshoot of 0.07 % and a settling time of 0.65 s. For an electric power step response, the MPC controller yields an overshoot of 0.28 % and a settling time of 36.6 s. These control performances are significantly better than those achieved by the PID controller. In power tracking control, the MPC controller demonstrates virtually error-free tracking performance. When external reactivity disturbances are introduced, the MPC controller outperforms the PID controller in swiftly compensating for the disturbances, maintaining the system's operating state at the desired level. [ABSTRACT FROM AUTHOR]

A deterministic neutron transport solver with SP 3 method based on OpenFOAM has been developed and validated in this paper. The physical models, numerical methods and schemes, boundary conditions, and solving strategy are discussed in detail. The approximate Marshak boundary condition of this solver is validated by the ISSA benchmark problem. Pin-by-pin lattice and simplified core neutron transport benchmark problems are introduced to make the validation of the solver. Numerical results show that SP 3 solver developed based on OpenFOAM has the promising prediction accuracy of k e f f , neutron scalar flux, fission power, and control rod worth. Parallel tests of SP 3 solver are also carried out under the million scale meshes, and the solver has an acceptable acceleration ratio and efficiency. No new computational methods or numerical schemes are proposed in this paper. Instead, the main new works in this paper are our detailed and quantified numerical validations of serval international neutron transport benchmark problems for SP 3 solver based on OpenFOAM. The capability of OpenFOAM in performing SP 3 calculations has been further tested and validated in this paper, which gives additional confidence that OpenFOAM can effectively be used for SP 3 analysis. • Detailed and quantified numerical validations of SP 3 solver based on OpenFOAM. • This study shows the accuracy of FV-based SP 3 method in specific benchmarks. • Additional confidence of SP 3 analysis based on OpenFOAM has been enhanced. [ABSTRACT FROM AUTHOR]

• Hydrogen dissociation in NTP reactors was analyzed by the numerical method. • Thermophysical properties of H-H 2 were established and validated from 200 to 3000 K. • Degree of dissociation was small and x H,out was just 0.456% in the NRX-A6 reactor. • Power density was one of the most significant factors for hydrogen dissociation. Nuclear thermal propulsion utilizes the nuclear reactor rather than the combustion chamber to yield thermal energy. Propellant hydrogen could dissociate in the high-temperature reactor, which has an important effect on thermal hydraulic performance of the reactor. In this study, a one-dimensional steady-state analysis code has been developed for studying the behavior of hydrogen flowing through the high-temperature coolant channel. Thermal dissociation and real gas thermophysical property models of hydrogen were proposed and considered in the calculation models. It was found that the model validation deviations of thermophysical properties were within ± 5% in the range of 200 ∼ 3000 K and 0.01 ∼ 10.0 MPa. Developed models were reliable and accurate with validation. Thermal-hydraulic behaviors of hydrogen in NRX-A6 reactor channel were analyzed. When dissociation occurred, the variation of properties was larger than those without dissociation, which enhanced heat transfer. The degree of dissociation was small and x H, out was just 0.456% under the design condition. The power density was the most significant influence factor, especially under the high power density. x H, out was 3.2 times than that of design condition as the power density grew 20%. This study can provide an approach to study hydrogen dissociation phenomena in nuclear thermal propulsion reactors. [ABSTRACT FROM AUTHOR]

In this study, a two-fluid model coupled with wall boiling model was adopted to investigate two-phase forced convection subcooled boiling process in a vertical heated pipe based on FLUENT 14.5. The influences of turbulence models ( k – ε models and k – w models) on flow and heat transfer in pipe were studied. The performances of different turbulent models, wall functions and two-phase turbulence treatments were analyzed based on four sets of grids by comparing the calculated results with Bartolemei experimental data. The results of k – ε models showed better agreement with experimental data than these of k – w models. The Dispersed and Per Phase treatments for two-phase turbulence parameters performed no better than the Mixture treatment. The enhanced wall function could not cope the situation for grid with near-wall Y -plus near to 1. k – ε models with appropriate mesh size performed excellent in predicting subcooled boiling in vertical channel. This work can be referred for choosing turbulence model when analyzing subcooled boiling by CFD methodology. [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]

In this paper, the supercritical-water cooled solid breeder test blanket module (SWCB TBM), using the supercritical water as the coolant, Li 4 SiO 4 lithium ceramic pebbles as a breeder, and beryllium pebbles as a neutron multiplier, was designed and analyzed for ITER. The results of neutronics, thermo-hydraulic and thermo-mechanical analysis are presented for the SWCB TBM. Neutronics calculations show that the proposed TBM has high tritium breeding ratio and power density. The tritium breeding ratio (TBR) of the proposed design is 1.17, which is greater than that of 1.15 required for tritium self-sufficiency. The thermo-hydraulic calculation proved that the TBM components can be effectively cooled to the allowable temperature with the temperature of outlet reaching 500 °C. According to thermo-mechanics calculation results, the first wall with the width of 17 mm is safe and the deformation of first wall is far below the limited value. All the results showed that the current TBM design was reasonable under the ITER normal condition. [ABSTRACT FROM AUTHOR]

Two-fluid model coupled with improved RPI wall boiling model (CHF model) was employed to investigate the departure from nuclear boiling (DNB) phenomena in vertical pipes under extremely high heat flux by using FLUENT 14.5 code. The distributions of temperature before and after DNB and the critical heat fluxes (CHF) at corresponding DNB points were obtained. The maximum deviation of predicted CHFs with experimental data was less than 15.0%. The results proved that two-fluid model coupled with CHF model was qualified to predict CHF accurately. This work can be referred when one predicts CHF in complicated geometry by CFD methodology. [ABSTRACT FROM AUTHOR]

The Modular Accident Analysis Program version 5 (MAAP5) is a computer code that can simulate the response of light water reactor power plants during severe accident sequences. The present work aims to simulate the severe accident of a typical Chinese pressure water reactor (PWR) with MAAP5. The pressurizer safety valve stuck-open accident is essentially a small break loss-of-coolant accident (SBLOCA), which becomes one of the major concerns on core melt initiating events of the PWR. Six cases with different assumptions in the pressurizer (PZR) safety valves (SVs) stuck-open accident stuck open accident were analyzed for comparison. The results of first three cases show that the severe accident sequence is correlated with the number of the stuck open valve. The primary system depressurized faster in a more SVs stuck open case, and the consequences in which is hence slighter. The remaining 3 cases along with the case 2 were then analyzed to study the effect of operator intervention to the accident. The results show that the auxiliary feed water (AFW) is effective to delay the core degradation and hence delayed the finally system recovery. The high pressure injection (HPI) operation and manually opening the steam generator (SG) SVs are effective to mitigate this kind of severe accident. The results are meaningful and significant for comprehending the detailed process of PWR severe accident, which is the basic standard for establishing the severe accident management guidelines. [ABSTRACT FROM AUTHOR]

Highlights: [•] A transient analysis code is developed to simulate the supercritical blowdown. [•] Models are established for the supercritical region and subcritical region. [•] The code is employed to calculate the supercritical blowdown for different fluids. [•] The code was verified by the blowdown experiment of supercritical CO2. [•] The blowdown of supercritical water was investigated in detail with the code. [Copyright &y& Elsevier]

Abstract: Using the MELCOR code, we simulated and analyzed a severe accident at a Chinese pressurized reactor 1000-MW (CPR1000) power plant caused by station blackout (SBO) with failure of the steam generator (SG) safety relief valve (SRV). The CPR1000 response and results for three different scenarios were analyzed: (i) seal leakage and an auxiliary feed water (AFW) supply; (ii) no seal leakage or AFW supply; and (iii) seal leakage but no AFW supply. The results for the three scenarios are compared with those for a simple SBO accident. According to our calculations, the SG SRV stuck in the open position would greatly accelerate the sequence for a severe accident. For an SBO accident with the SRV stuck open without seal leakage or an AFW supply, the pressure vessel would fail at 9576 s and the containment system would fail at 124,000 s. If AFW is supplied, pressure vessel failure would be delayed nearly 30000 s and containment failure would delay at least 50000 s. When seal leakage exists, pressure vessel failure is delayed about 50 s and containment failure time would delay about 30000 s. The results will be useful in gaining an insight into the detailed processes involved and establishing management guidelines for a CPR1000 severe accident. [Copyright &y& Elsevier]

Highlights: [•] We study the SCWR steam cycle with a two-stage reheating and an eight-stage regenerative system. [•] Average Distribution Method is used to study the parameter relationships. [•] High coolant enthalpy rise can reduce coolant flow rate and coolant load. [•] 280°C will be a proper temperature selected for the core inlet. [•] The thermal efficiency is larger than 45% when the core outlet temperature is 566°C. [ABSTRACT FROM AUTHOR]

Highlights: [•] Spraying characteristics in the pressurizer of Pressurized Water Reactor were tested. [•] Spray droplet collector was fabricated to obtain a 2-D spray flux distribution graph. [•] A photography chamber was specially implemented to measure the spray drop size. [•] An exponent was proposed for the discharge coefficient of the pressure-swirl nozzle. [•] A better empirical method for Probability Density Function of drop size was proposed. [Copyright &y& Elsevier]

Abstract: The T-junctions are widely used in industry, such as the joint between hot leg and Automatic Depressurization System Stage Four (ADS-4)/Passive Residual Heat Removal (PRHR)/surge pipeline in AP1000 plant, header-feeder geometry of CANDU reactor, complex piping systems in the water and petroleum processing and so on. In the nuclear plant accident analysis, the T-junction entrainment is fairly closely linked with the water volume of reactor and reactor safety. Many researchers have conducted the experimental and theoretical research on the generation process, mechanism, influencing factors and modeling of entrainment. Research directions of entrainment at T-junction are multifarious, so a review work is necessary. This paper describes a comprehensive literature review of entrainment at T-junction. Based on different geometric structures of T-junction, detailed summarizations and reviews on the present entrainment phenomena, entrainment models and scaling analysis are performed. Conclusions and further research directions on entrainment at T-junction are proposed. [Copyright &y& Elsevier]

• The influence of geometric parameters including twist pitch and cross section sizes on friction factor and heat transfer coefficient is studied in the helical cruciform single rod. • Geometric dimensionless numbers are proposed to fit the friction factor and heat transfer coefficient of helical cruciform. • The proposed empirical correlations of frction factor and heat transfer coefficient are validated respectively based on numerical simulation. Helical cruciform fuel is a novel type of nuclear reactor fuel with self-supporting and large surface-volume ratio features, potential to increase the reactor power density. However, the shape of the fuel is so complicated that influence of geometric parameters on thermal hydraulic characteristics was not clearly researched. In this paper, geometric dimensionless numbers are proposed and validated based on the numerical simulation to modify friction factor and heat transfer coefficient correlations proposed through experiments of helical cruciform fuels. Twist pitch of the fuel has a great impact on flow resistance and heat transfer, resulting in the modified coefficient ranging from 0.94 to 1.04 for friction factor and from 0.022 to 0.024 for heat transfer coefficient respectively. Change in cross-section parameters has little influence on the fuel performance, leading to the unit cross-section modified factor for both friction factor and heat transfer coefficient. After fitting the correlations with the dimensionless number, several cases are established to validate the empirical correlations, with a maximum error below 10.0 % for friction factor equation and 8.5 % for heat transfer coefficient correlation when Reynolds number is between 2.7 × 104 and 2.7 × 105. This paper provides a guide for study of heat transfer and fluid flow in helical structures, and a reference for establishment of reactor system model using helical cruciform fuels. [ABSTRACT FROM AUTHOR]