Search

• Rolling motion significantly affects the bubble rise trajectory and rise velocity. • Effect of rolling increases with rolling amplitude and decreases with rolling period. • Effect of rolling force on bubble trajectory decreases with rise velocity. • Distance of vessel from rolling pivot has a significant effect on rolling forces on bubble. Purpose of this research is to study the effect of rolling motion on single gas bubble rising in stagnant liquid. The effect of rolling motion has been incorporated in the Navier-Stokes equations and simulations have been performed using MPS – MAFL technique in two dimensions. Some experiments have also been conducted to study the phenomenon for validation purposes. The bubble trajectory, rise velocity, aspect ratio and shapes obtained from simulations have been compared with experimental data. The numerical results have been found in good agreement with the experimental findings. Effect of various factors including rolling amplitude, rolling period, rolling phases, distance from the rolling pivot, liquid properties and bubble diameter have been studied on bubble trajectory and rise velocity under the rolling motion conditions. It has been found that effect of rolling conditions is enhanced as the rolling amplitude is increased or rolling period is decreased. [ABSTRACT FROM AUTHOR]

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

In this paper, a board type sodium cooled breed and burn reactor core is further designed and applied to perform minor actinide (MA) transmutation. MA is homogeneously loaded in all the fuel sub-assemblies with a weight fraction of 2.0 wt.%, 4.0 wt.%, 6.0 wt.%, 8.0 wt.%, 10.0 wt.% and 12.0 wt.%, respectively. The transmutation efficiency, transmutation amount, power density distribution, neutron fluence distribution and neutronic safety parameters, such as reactivity, Doppler feedback, void worth and delayed neutron fraction, are compared under different MA weight fraction. Neutronics and depletion calculations are performed based on the self-developed MCNP–ORIGEN coupled code with the ENDF/B-VII data library. In the breed and burn reactor core, a number of breeding sub-assemblies are arranged in the inner core in a board type way (scatter load) to breed, and a number of absorbing sub-assemblies are arranged in the inner side of the outer core to absorb neutrons and reduce power density in this area. All the fuel sub-assemblies (ignition and breeding sub-assemblies) are shuffled from outside in. The core reached asymptotically steady state after about 22 years, and the average and maximum discharged burn-up were about 17.0% and 35.3%, respectively. The transmutation amount increased linearly with the MA weight fraction, while the transmutation rate parabolically varied with the MA weight fraction. Power density in ignition sub-assembly positions increased with the MA weight fraction, while decreased in breeding sub-assembly positions. Neutron fluence decreased with the increase of MA weight fraction. Generally speaking, the core reactivity and void worth increased with the MA weight fraction, while the Doppler feedback and the effective delayed neutron fraction decreased with the increase of the MA weight fraction. [ABSTRACT FROM AUTHOR]

Highlights: [•] MCNP and ORIGEN are coupled to perform nuclides depletion and decay calculation. [•] Coupled system MCORE uses “modified predictor corrector” approach. [•] MCORE can use different depletion schemes and simulate fuel shuffling. [•] MCORE is assessed by a “VVER-1000 LEU Assembly Computational Benchmark”. [•] MCORE is also assessed by a fast reactor benchmark problem. [Copyright &y& Elsevier]

Abstract: Emergency Passive Safety System (EPSS) is an innovative design to improve reliability of nuclear power plants. In this work, the EPSS consists of secondary passive residual heat removal system (PRHRS) and the reactor Core Makeup Tank (CMT) system. The PRHRS, which has been studied in our previous paper, can effectively remove the core residual heat and passively improve the inherent safety by passive methods. The designed CMT, representing the safety improvement for CPR1000, is used to inject cool boron-containing water into the primary system during the loss of coolant accident. In this study, the behaviors of EPSS and transient characteristics of the primary loop system during the Steam Generator Tube Rupture (SGTR) accident are investigated using the nuclear reactor thermal hydraulic code RELAP5/MOD3.4. The results show that the designed CMT can protect the reactor primary loop from boiling and maintain primary loop coolant in single phase state. Both PRHRS and CMT operation ensures reactor safety during the SGTR accident. Results reported in this paper show that the designed CMT is a further safety improvement for CPR1000. [Copyright &y& Elsevier]

Abstract: The comparative studies of transient flow characteristics and the consequent safety performance for SCWRs with different flow path designs have been performed. Three different SCWR concepts SCWR-M, American SCWR and SSCWR are respectively selected as the representations of multi-pass core, one-pass core with water rods and one-pass core without water rods. The comparative analyses of loss of flow accident (LOFA) are carried out for the three SCWR concepts by TACOS code. The flow rate and cladding temperature behaviors are compared to each other. The results show that the SCWR-M and the American SCWR concepts have more complicated transient flow characteristics than the SSCWR. Under the assumed condition in the work, inverse flow occurs both in the SCWR-M and the American SCWR following the LOFA. The evaluation of the effect of AFS flow rate indicates that the flow path design of the SCWR-M would require larger injection flow rate than the other two concepts to mitigate the LOFA. [Copyright &y& Elsevier]

Abstract: The reliable prediction of nuclear fuel rod behavior is of great importance for safety evaluation of nuclear reactors. In the present study, a thermo-mechanical coupling code FROBA (Fuel ROd Behavior Analysis) has been independently developed with consideration of irradiation and burnup effects. The thermodynamic, geometrical and mechanical behaviors have been predicted and were compared with the results obtained by Idaho National Laboratory to validate the reliability and accuracy of the FROBA code. The validated code was applied to analyze the fuel behavior of AP1000 at different burnup levels. The thermal results show that the predicted peak fuel temperature experiences three stages in the fuel lifetime. The mechanical results indicate that hoop strain at high power is greater than that at low power, which means that gap closure phenomenon will occur earlier at high power rates. The maximum cladding stress meets the requirement of yield strength limitation in the entire fuel lifetime. All results show that there are enough safety margins for fuel rod behavior of AP1000 at rated operation conditions. The FROBA code is expected to be applied to deal with more complicated fuel rod scenarios after some modifications. [Copyright &y& Elsevier]

Abstract: Flow instability was studied for supercritical water-cooled reactor CSR1000 based on frequency domain method in this paper. A code named FREDO-CSR1000 (Frequency Domain Analysis of CSR1000) has been developed. The analysis of flow instability for CSR1000 both in average channel and hot channel within rated power and flow has been performed with the FREDO-CSR1000 code. The key parameters including decay ratio, the maximum cladding surface temperature and the proper orifice pressure drop coefficient were obtained as well. The calculation results indicated that the decay ratio of the first flow path varies with power and flow monotonically. However, the decay ratio of the second flow path ascends first and then descends, the trend of which is fluctuant because of the simultaneous influence of a multitude of variables. Besides, it is found that the location where the flow instability happened is directly determined by the point at which the pseudo-critical temperature reached. Moreover, the neutral stability boundary for hot channel of CSR1000 was obtained so that the threshold of flow instability can be predicted. [Copyright &y& Elsevier]

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

Abstract: In this study, the method of characteristic line (MOC) was adopted to evaluate the valve-induced water hammer phenomena in a parallel pumps feedwater system (PPFS) during the alternate startup process of parallel pumps. Based on closed physical and mathematical equations supplied with reasonable boundary conditions, a code was developed to compute the transient phenomena including the pressure wave vibration, local flow velocity and slamming of the check valve disc, etc. Some interesting results were obtained and it was shown that severe slamming between the valve disc and valve seat occurred during the alternate startup of parallel pumps. The induced maximum pressure vibration amplitude is up to 5.0MPa, which occurs under the high–high speed startup condition. The scheme of appending a damping torque with the check valve disc was also numerically performed to eliminate the water hammer for the optimum design purpose. The adoption of damping torque slows down the closing speed of the check valve and has been approved to be an effective approach. This work is expected to be instructive for the optimum design of the PPFS in NPPs so as to mitigate the potential damage caused by valve-induced water hammer. [Copyright &y& Elsevier]

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

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

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

• A three-field mechanistic model is developed for dryout prediction in annular flow. • The model is verified by experimental data of the circular and rectangular channel with relatively large parameter range. • Some key constitutive relations in the model are discussed for the better prediction. • The model is effective on prediction of dryout in the rod bundles by coupling with subchannel code. Critical heat flux (CHF) is one of the most important thermal criteria for nuclear power plants. It has traditionally been evaluated using look-up tables or empirical correlations. In this paper an annular film dryout (AFD) mechanistic model has been developed based on the interaction of three fields in annular flow region, i.e. the liquid film, entrained droplets and vapor core. The model describes the mass and momentum conservation equations of three fields together with a series of constitutive relations. The effect of some constitutive correlations (the entrainment and deposition of droplets and the onset of annular flow) on the prediction accuracy of the model is studied. The results are compared with CHF experimental data in flow boiling. Fairly good agreements are observed for the CHF in circular tubes with uniform and axially non-uniform heating, as well as rectangular channels with uniform heating. Through coupling with the subchannel analysis method, the model is used to predict the dryout-type CHF in the rod bundles with a good precision. [ABSTRACT FROM AUTHOR]

• Two quick-analysis models are investigated and compared in theory and application. • The simulations verify and compare the two models via the typical sodium heat pipe frozen startup experiment. • Inherent defects and targeted model improvements of the two models are proposed. The alkali-metal high-temperature Heat Pipe (HP) is the common component connecting the reactor core and energy conversion device in Heat Pipe cooled Reactor (HPR). Due to the particularity of the working medium, the frozen startup of HP makes a great influence on the startup process of HPR system. In this paper, two quick-analysis models, the Improved Thermal Resistance Network (ITRN) method and the Pseudo Wick Thermal Conductivity (PWTC) method have been investigated and compared. The simulations of a sodium HP in a typical experiment are carried out by two methods in COMSOL. For simulation speed, the convergence time of a single HP using the quick-analysis models is about tens to hundreds of seconds depending on the computing equipment. The convergence of PWTC is 26.7% faster than that of ITRN. For simulation accuracy, except for the inevitable error near the flat-front, the relative error of ITRN is less than 6.94% and the relative error of PWTC is less than 8.24%. The maximum effective error of ITRN method is 34.84 K, while that of PTWC method is 58.21 K at the initial stage of startup. The T Transition = 651 K in PWTC is lower than the T exp. = 710 K in the experiment and the T Tr = 701 K in ITRN, which affects the accuracy at the initial stage of startup. The error will decrease gradually in the thawing front advancing stage. For simulation analysis, some inherent defects and targeted model improvements of the two models are proposed. Through the detailed comparative study in this work, researchers can easily apply the appropriate model according to the specific requirements and avoid the defects during the simulation process. [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]

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]

• 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]

• 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]

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]

• 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]

• 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]

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]

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]

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]

• 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]

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

• A conceptual design of FuSTAR is developed, and validated by neutronics and thermal-hydraulics analysis. • Detailed core design and neutronics analysis with near-critical configuration and burnup calculation are carried out. • Detailed thermal-hydraulics analysis with turbulent flow and heat transfer is carried out. The fluoride-salt-cooled high-temperature advanced reactor (FuSTAR) is an advanced small modular reactor, which aims to the electric supplement and clean heat production to western China. In order to have good economy and safety, FuSTAR needs to have the characteristics of small size, light weight, deep burnup, and inherent safety. A conceptual design of FuSTAR is proposed in this study with neutronics and thermal-hydraulics assessments. In this study, its concept is established with two-region fuel to flatten core power, and the neutron spectrum of the core is in the thermal spectrum to improve the utilization of thermal neutrons and reduce fuel inventory. Neutronics and thermal-hydraulics analyses are also carried out, and established the core layout and material ratio under the near-critical condition. Burnup depth of the fuel, power distribution, neutron flux distribution, flow and heat transfer in rod-bundles channel, and 1/6 core porous media analysis are also investigated. The parameters of the designed core are evaluated quantitatively from various aspects. The results show that the core design in this study has sufficient shutdown depth, the equilibrium cycle is 3.24 years, and the discharge burnup depth is 106 GWd/MtU. The helical cruciform fuel (HCF) adopted by this reactor has a better flow and heat transfer capacity than cylindrical-type fuel at the inner channel. The analysis of 1/6 core porous media shows that the peak fuel temperature under normal operating condition is 1009.3 K, which is far below the limit of 1573 K. Relevant design and analysis results can provide the reference for the subsequent small modular FHR design and construction. [ABSTRACT FROM AUTHOR]

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

• Subchannel code for the thermal hydraulic analysis of liquid metal cooled reactor. • Simulation of transition from forced to natural circulation by subchannel code. • Effect of non-uniform heating on the thermal field. • Validation of YAOGUANG code under steady and transient conditions for LBE coolant. Based on the experimental campaign performed on the NACIE-UP (NAtural Circulation Experiment-Upgraded) facility, the benchmark activity was proposed by ENEA to qualify and validate the numerical tools for the Heavy Liquid Metal (HLM) system. Xi'an Jiaotong University participated in the benchmark exercise of subchannel validation and the self-developed subchannel code YAOGUANG was applied to the thermal hydraulic analysis. The performed two tests with different power distributions on the Fuel Pin bundle Simulator (FPS) were simulated, characterized by transition from forced to natural circulation. The applicability of this code to the stable state and mass flow rate transient was evaluated by comparing the numerical outcomes with the experimental results. Then the effect of the non-uniform heating on the thermal field was analyzed. The results indicate that this subchannel code has good accuracy in simulating both steady state and transient conditions with uniform power distribution. For the non-uniform case, the subchannel calculations can capture the general trend of temperature distributions, with a larger simulation deviation observed. This discrepancy is mainly due to the limitations of existing models adopted because they didn't consider the impacts of non-uniform conditions on the flow and heat transfer in rod bundles. Finally, the influence of non-uniform heating is mainly manifested as the increase of axial and radial temperature gradients and it helps to flatten the radial thermal field to some degree. [ABSTRACT FROM AUTHOR]

Abstract: The steam generator (SG) secondary water-cooled and air-cooled passive residual heat removal systems (PRHRSs) are proposed and designed for 300MW Pressurized Water Reactor (PWR). The RELAP5/MOD3.4 code is utilized to study the behavior of the PRHRSs and transient characteristics of primary loop system during Feed-water Line Break (FLB) accident. The characteristic comparison between water-cooled and air-cooled PRHRSs is also conducted in this study. The results show that both water-cooled and air-cooled PRHRSs can establish stable natural circulations in PRHRS loops, which realize the effective core decay heat removal from primary loop. Results illustrate that both water-cooled PRHRS and air-cooled PRHRS designed in the study have great significance for improving the inherent safety of 300MW nuclear power plant. However, the water-cooled PRHRS heat transfer ability is stronger in the initial time, while it becomes weaker as the system tank water evaporates in the later stage. In contrast, the air-cooled PRHRS, ensuring the main reactor operation parameters to decrease to a more safety value, is more effective than water-cooled PRHRS for long term cooling. [Copyright &y& Elsevier]

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

• A sub-channel code which is capable of modelling the flow blockage has been developed and verified. • The results of flow blockage of the inner channel and out channel has been obtained. • The effect of the flow blockage on the heat split and DNBR has been discussed. A sub-channel code capable of modelling the flow blockage of the annular fuel has been developed and verified by the experiment. A fuel conduction model has been developed for the blockage to consider the effect of the blockage on the temperature distribution of the fuel. Blockage at different position has been modeled and the middle blockage is selected as the calculation object. Blockage of inner and outer channel with different blockage rate has been calculated and the results of fuel temperature distribution, the heat split, the mass flow rate and the DNBR distribution has been obtained. Based on the analysis of the results, the blockage of inner and outer channel will have different effects on mass flow and heat split of the fuel, and the blockage of inner channel has a greater impact on DNBR and fuel temperature than that of outer channel. [ABSTRACT FROM AUTHOR]

Numerical simulation analyses of LIFUS5/MOD2 experiment facility for hyperbaric steam-water jet inside liquid lead bismuth eutectic (LBE) were conducted using computational fluid dynamics (CFD). Alchagirov surface tension correlation and Lee evaporation model were applied in Eulerian-Eulerian two-phase flow model. Benchmark study was performed by comparing the CFD results against experimental data. Four thermal-hydraulics phenomena of sensitivity analysis were established under three extended conditions. These include 1) argon zone compression risk, 2) pressure rising and initial pressure peak, 3) argon entrained by LBE fluctuation, and 4) LBE level rising and steam cavity diffusion path. Overall, CFD models have a great advantage simulating pressure change and pressure wave transfer with 0.37% maximum error. LBE level rising and the radial and even diffusion of steam cavity should happen simultaneously under steam generator tube rupture accident (SGTR). This study provides the engineering basis for Lead-cooled Fast Reactor (LFR) design to improve its safety margin. [ABSTRACT FROM AUTHOR]

• The visual entrainment experiments in the vertical tee branch are carried out, and the test data and videos are obtained. • The effects of branch Angle, air intake mode and working medium on entrainment are revealed. • The new OLE model and entrainment rate model are proposed, which can be used for nuclear reactor safety analysis. The tee branch structure is broadly used in the nuclear power systems, and liquid entrainment in the tee branch has been studied in depth. However, most of the existing research focuses on the vertical tee branch, while the study about the entrainment phenomenon in the inclined tee branch hasn't formed mature contents. This research aims to analyze the liquid entrainment in the inclined up tee branch. The liquid level measurement method of the Automatic Depressurization and Entrainment TEst Loop (ADETEL) was improved and supplementary experiments were conducted with single-end gas inlet of vertical tee branch. In addition, the test loop was modified to conduct visualization experiments on liquid entrainment experiments of the tee branch with a 45 degree angle. The experimental result reveals that the liquid entrainment phenomenon in the inclined tee branch is significantly different from that in the vertical tee branch. Based on the experimental phenomena and data, a new OLE (Onset of Liquid Entrainment) model is proposed which is well fitted with the existing data. Then, an entrainment rate model is proposed by nondimensionalizing the key parameters affecting the steady-state entrainment, with the error within ± 15% of the experimental data. [ABSTRACT FROM AUTHOR]

• Validation of THACS and OpenMC for SFR has been enhanced by simulating FFTF LOFWST Test #13. • A new modeling method multi-parallel channel plus local subchannel model is proposed for reactor core. • Important local phenomenon like inter wrapper flow, thermal stratification is analyzed by the specific model. In order to support collaborative efforts within international partnerships on the development of sodium-cooled fast reactor (SFR), the International Atomic Energy Agency (IAEA) proposes a coordinated research project (CRP) about the unprotected loss-of-flow-without-scram test (LOSWOS) performed at Fast Flux Test Facility (FFTF) in 2018, which could validate simulation tool and numerical models for SFR. As one of the participants, the Nuclear THermal-hydraulic Laboratory at Xi'an Jiaotong University (NuTHeL) participates with the monte carol code OpenMC and system thermal hydraulic code THACS. This paperdeal with the standalone neutronic modeling by OpenMC and thermal hydraulic modelling by THACS. For neutronic part, the axial homogenous model was modeled to analyze the reactivity feedback coefficients and steady radial power distribution. For thermal hydraulic part, the system model, including 2D CFD outlet plenum model, subchannel model and inter-wrapper flow model, was modeled to analyze both system response and local phenomena during the LOSWOS. In general, both neutronic results and thermal hydraulic results agreed well with other participants or experiments, the reactivity feedback of GEM could provide major negative feedback, which could be used as a novel passive shutdown device. However, the remaining discrepancies were the underestimated peak temperature of PIOTA at the begin of transients, it may be caused by unknow heat transfer coefficients around hexagon tube, which has no correlation and should be further investigated by experiments. In addition, when considering IWF model, the maximum temperature of the outlet for the active fuel region was 10 K larger than the subassemblie' outlet in the PIOTA, this phenomenon should be carefully treated in loss of flow accidents of SFR. [ABSTRACT FROM AUTHOR]

• In order to obtain the behavior of axial flow field evolution under the effect of intrinsic structure on fluid excitation, we focus on the fluid excitation of fuel rod with spacer grids in axial flow by three-dimensional numerical methods. • The effects of the clamping structures and mixing vanes are investigated separately, and their effects on the excitation force are analyzed by the simulation results. • The mechanism and causes of fluid excitation force generation are described. • The results of the excitation force and the frequency spectrum is analyzed. Grids, supporting fuel rods, consist of dimples and mixing vanes. These complex structures cause turbulent flow field with complex vortex. The vibration of fuel rod by the high-velocity coolant may lead to grid to rod fretting (GTRF). In order to obtain the behavior of the fretting wear, it is crucial to investigate the induction mechanism of fuel rod by the flow field evolvement. In this paper, the effects of two structures with different geometric were discussed by numerical simulation. The generation and evolution of the fluid forces are analyzed in detail. The results show that the fluid forces mainly originate from the pressure asymmetry. Moreover, the clamping structures play a dominant role in both excitation and fluctuation of the fluid forces. The study obtained the effect of intrinsic structure on fluid excitation, and it can provide a basis for the fluid excitation boundary of the fretting wear of fuel rods. [ABSTRACT FROM AUTHOR]

• CFD method is becoming more important in reactor thermal hydraulic analysis. • The latest progress of CFD applications in PWR is reviewed. • The major challenges and promising directions of CFD are proposed. CFD method has the potential to simulate the detailed three-dimensional flow and heat transfer features in nuclear reactors, and is promising to play a more important role in the future reactor design and thermal hydraulics analysis. In recent years, series of research achievements in nuclear engineering based on the CFD method, including the single-phase application, two-phase model development and multi-scale and multi-physics coupling, are accomplished around the world. XJTU-NuTheL has also been committed to the development and application of high fidelity thermal–hydraulic models using CFD method. The three-dimensional CFD models of key equipments in nuclear power plants, such as RPV, SG, valves, T-junctions and passive residual heat exchanger, are developed. The mathematical models of complicate two-phase boiling phenomena and thermal hydraulic features under the motion conditions are established. In addition, with the high fidelity simulation requirement of the whole reactor system, the nuclear reactor multi-scale and multi-physics coupling platforms are developed on the basis of CFD codes. In this paper, the latest progress of nuclear reactor thermal–hydraulic research using CFD method is outlined, especially at XJTU-NuTheL. The major challenges and promising directions of CFD method in the nuclear reactor engineering are proposed, which would be beneficial for the promotion of its further applications. [ABSTRACT FROM AUTHOR]

• Transient tritium transport characteristics in the TMSR-SF have been analyzed. • Tritium permeation rates under the URIA and UOC are lower than steady condition. • Extra tritium control facility under these accident conditions is unnecessary. Tritium management is a vital factor that needs to be considered before a molten salt reactor (MSR) or fluoride-salt-cooled high-temperature reactor (FHR) is constructed. The analysis of tritium transport characteristics under steady and transient conditions is essential in the conceptual design of MSRs and FHRs. In this study, tritium transport characteristics analysis is conducted for a fluoride-salt-cooled pebble bed reactor, i.e., thorium molten salt reactor with solid fuel (TMSR-SF) under accident conditions. The tritium transport characteristics analysis (TAPAS) code is improved with a power model and heat-and-mass transfer model to adapt to TMSR-SF. The transient responses of the tritium transport characteristics of TMSR-SF under the unprotected reactivity insertion accident (URIA) and unprotected overcooling accident (UOC) are determined. It is found that the amount of tritium fluoride (TF) adsorbed on graphite varies, indicating the dynamic equilibrium between the TF adsorption and desorption on graphite. Due to the nuclear density variations of Li-6 and Be-9 in the fluoride salt under the accident conditions, the tritium production rate increases and decreases under the UOC and URIA conditions, respectively. Generally, under these two accident conditions, the tritium permeation rates are lower than that of the normal operation condition. In these cases, it is unnecessary to develop extra safety facilities for tritium control. This study contributes to the tritium management of TMSR-SF. [ABSTRACT FROM AUTHOR]

Thermal characteristics and mechanical analysis are particularly essential in the performance and safety analysis of metallic fuel for Liquid Metal Fast Reactor (LMFR). Therefore, the development of advanced thermo-mechanical analysis code is significant for the safety design of metallic fuels. In order to meet the future demand for metallic fuel application in China, the FRAC code was developed to analyze the metallic fuel performance during steady-state conditions. The code was benchmarked against the open-literature EBR-II experimental database. In comparison with the experimental data given, the code calculated Zr mole fraction, fuel temperature distribution, fission gas release and clad strain distribution are in good agreement with the data, which indicates that the code is reliable and can be applied into safety analysis of metal fuel design in China. The overall code assessment shows that the prediction error is within acceptable range. [ABSTRACT FROM AUTHOR]

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

• Steam condensation on a downward-facing plate with inorganic zinc coating. • The effects of key parameters on condensation heat transfer. • An empirical correlation for condensation heat transfer coefficient prediction. Steam condensation on the containment vessel is a significant cooling strategy to cope with the design basis accidents in the third-generation reactors with passive safety features. Aiming to evaluate the heat removal capacity of steam condensation, a number of condensation experiments have been conducted on a horizontal plate with the same surface condition as CAP1400 containment vessel. The effects of surface sub-cooling, air concentration, bulk pressure and different steam injecting models on condensation heat transfer have been experimentally investigated. In the experiments, the condensation heat transfer coefficients (within the range of 77–1297 W/m2 K) on the downward-facing plate are successfully obtained under the conditions of bulk pressure ranging from 0.154 to 0.607 MPa, air molar fraction ranging from 1.1% to 82.9% and surface sub-cooling ranging from 12.6 to 49.3 °C. In addition, an empirical correlation for heat transfer prediction has been developed based on the experimental data in this work. Besides, the experimental data have also been compared with four existing empirical correlations proposed by Uchida, Tagami, Dehbi and Liu, respectively. The comparing results show that most of condensation heat transfer coefficient data on the test plate could be overestimated by these four correlations. [ABSTRACT FROM AUTHOR]

• A new boron diffusion model is proposed for the nuclear reactor core analysis. • The effect of turbulent diffusion is taken into consideration. • The model is implemented in the subchannel analysis code. • The CFD method is used to verify the new model. The diffusion characteristic of borated water in nuclear reactor core is important for the reactivity control during some possible reactivity transient scenarios. The non-uniform distributions of boron concentration increase the risk of reactor accidents and lead to some reactor safety problems. Hence it is necessary to develop an effective boron diffusion model into the reactor thermal-hydraulic analysis code. In this paper, a quasi-three dimensional boron tracking model, in which the turbulent diffusion effect was considered carefully, was proposed and successfully implemented in the subchannel analysis code. The reliable CFD method with k-omega turbulent model validated against International Standard Problem (ISP No. 43) experimental data was used to verify the feasibility of proposed boron tracking model. A typical Pressurized Water Reactor (PWR) 3 × 3 fuel bundle model was established and the borated water was injected into one specified channel and pure water into other channels. The distributions and variations of boron concentration along the flow direction were achieved and analyzed. Results indicate that the proposed boron tracking model could predict the boron diffusion process in the fuel rod bundles more accurately compared with the original model. This work provides an essential supplement for the nuclear reactor subchannel analysis method. [ABSTRACT FROM AUTHOR]

• The AP1000 adopts a new spiral-shaped pressurizer surge line design. • The thermal stratification phenomenon in AP1000 surge line is investigated. • AP1000 pressurizer surge line has smaller temperature difference with shorter period. • This work is meaningful for the AP1000 pressurizer surge line fatigue assessment. The surge line is implemented to connect the pressurizer with reactor cooling system in Pressurized Water Reactor (PWR). The integrity of surge line plays an important role for reactor safe operation. In the literatures, much work has been contributed to the thermal hydraulic characteristics and reliability research on traditional pressurizer surge line. For the advanced Generation-III PWR, the AP1000 adopts a new spiral-shaped pressurizer surge line concept. In this paper, the three-dimensional thermal hydraulic transient characteristics in AP1000 pressurizer surge line are studied and the formation mechanism of surge line thermal stratification phenomenon is investigated. The variation of maximum temperature difference with time in the whole surge line is achieved. Results show that the thermal stratification is becoming weaker along the flow direction and the most dangerous region is the entrance of horizontal position. The thermal stratification affected by the circumference effect is not significant during the surge line thermal fatigue analysis. In addition, through the comparison of temperature distribution in pressurizer surge line between the traditional PWR and AP1000, it can be seen that the AP1000 pressurizer surge line has smaller temperature difference at the corresponding position with shorter period. The new design concept of AP1000 surge line brings a great advantage to reduce the hazard causing by the thermal stratification. This work provides a valuable guideline for the AP1000 pressurizer surge line fatigue assessment. [ABSTRACT FROM AUTHOR]

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

This paper proposes an optimized design for the Heat Pipe ThermoElectric Generator (HPTEG) principle prototype based on the design of the NUSTER system. The numerical analysis platform has been established and verified. It is used to analyze the steady-state and transient (start-up and shutdown conditions) thermoelectric properties of the refined HPTEG prototype. The steady-state calculation shows that when the input power of the HPTEG prototype is 5500 W, the maximum temperature is about 987 K, and the thermoelectric conversion efficiency is about 10.64%. The startup calculation show that the heat pipe startup is completed at about 14400 s. The shutdown calculation shows that at 6% shutdown power, the system tends to be stable after about 20000 s, and the maximum temperature is about 716 K. This study provides data support for the subsequent design modification of the future HPTEG principle prototype and provides ideas for the 3D transient calculations of the solid-state heat pipe reactors. [ABSTRACT FROM AUTHOR]