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Reactor fuel's power distribution plays a vital role in designing the new generation thermionic Space Reactor Power Systems (SRPS). In this paper, the 1/12th SPACE-R's full reactor core was numerically analyzed with two kinds of different axial power distribution, to identify their impacts on thermal-hydraulic and thermoelectric characteristics. In the benchmark study, the maximum error between numerical results and existing data or design values ranged from 0.2 to 2.2%. Four main conclusions were obtained in the numerical analysis: a) The axial power distribution has less impact on coolant temperature. b) Axial power distribution influenced the emitter temperature distribution a lot, when the core power was cosine distributed, the maximum temperature of the emitter was 194 K higher than that of the uniform power distribution. c) Comparing to the cosine axial power distribution, the uniform axial power distribution would make the maximum temperature in each component of the reactor core much lower, reducing the requirements for core fuel material. d) Voltage and current distribution were similar to the axial electrode temperature distribution, and the axial power distribution has little effect on the output power.

A heat pipe residual heat removal system is proposed to be incorporated into the reactor driven subcritical (RDS) facility, which has been proposed by MIT Nuclear Reactor Laboratory for testing and demonstrating the Fluoride-salt-cooled High-temperature Reactor (FHR). It aims to reduce the risk of the system operation after the shutdown of the facility. One of the main components of the system is an air-cooled heat pipe heat exchanger. The alkali-metal high-temperature heat pipe was designed to meet the operation temperature and residual heat removal requirement of the facility. The heat pipe model developed in the previous work was adopted to simulate the designed heat pipe and assess the heat transport capability. 3D numerical simulation of the subcritical facility active zone was performed by the commercial CFD software STAR CCM + to investigate the operation characteristics of this proposed system. The thermal resistance network of the heat pipe was built and incorporated into the CFD model. The nominal condition, partial loss of air flow accident and partial heat pipe failure accident were simulated and analyzed. The results show that the residual heat removal system can provide sufficient cooling of the subcritical facility with a remarkable safety margin. The heat pipe can work under the recommended operation temperature range and the heat flux is below all thermal limits. The facility peak temperature is also lower than the safety limits.

Flow blockage of the fuel assembly in the lead-based fast reactor (LFR) may produce critical local spots, which will result in cladding failure and threaten reactor safety. In this study, the flow blockage characteristics were analyzed with the sub-channel analysis method, and the circumferentially-varied method was employed for considering the non-uniform distribution of circumferential temperature. The developed sub-channel analysis code SACOS-PB was validated by a heat transfer experiment in a blocked 19-rod bundle cooled by lead-bismuth eutectic. The deviations between the predicted coolant temperature and experimental values are within ±5%, including small and large flow blockage scenarios. And the temperature distributions of the fuel rod could be better simulated by the circumferentially-varied method for the small blockage scenario. Based on the validated code, the analysis of blockage characteristics was conducted. It could be seen from the temperature and flow distributions that a large blockage accident is more destructive compared with a small one. The sensitivity analysis shows that the closer the blockage location is to the exit, the more dangerous the accident is. Similarly, a larger blockage length will lead to a more serious case. And a higher exit temperature will be generated resulting from a higher peak coolant temperature of the blocked region. This work could provide a reference for the future design and development of the LFR.

The Free-Piston Stirling engine (FPSE) is of interest for many research in aerospace due to its advantages of long operating life, higher efficiency, and zero maintenance. In this study, a 1-kW FPSE was proposed by analyzing the requirements of Space Reactor Power Systems (SRPS), of which performance was evaluated by developing a code through the Simple Analysis Method. The results of SAM showed that the critical parameters of FPSE could satisfy the designed requirements. The heater of the FPSE was designed with the copper rectangular fins to enhance heat transfer, and the parametric study of the heater was performed with Computational Fluid Dynamics (CFD) software STAR-CCM+. The Performance Evaluation Criteria (PEC) was used to evaluate the heat transfer enhancement of the fins in the heater. The numerical results of the CFD program showed that pressure drop and Nusselt number ratio had a linear growth with the height of fins, and PEC number decreased as the height of fins increased, and the optimum height of the fin was set as 4 mm according to the minimum heat exchange surface area. This paper can provide theoretical supports for the design and numerical analysis of an FPSE for SRPSs.

One-dimensional oxygen transport relation is indispensable to study the oxygen distribution in the LBE-cooled system with an oxygen control device. In this paper, a numerical research is carried out to study the oxygen transport characteristics in a gas-phase oxygen control device, including the static case and dynamic case. The model of static oxygen control is based on the two-phase VOF model and the results agree well with the theoretical expectation. The model of dynamic oxygen control is simplified and the gas-liquid interface is treated as a free surface boundary with a constant oxygen concentration. The influences of the inlet and interface oxygen concentration, mass flow rate, temperature, and the inlet pipe location on the mass transfer characteristics are discussed. Based on the results, an oxygen mass transport relation considering the temperature dependence and velocity dependence separately is obtained. The relation can be used in a one-dimensional system analysis code to predict the oxygen provided by the oxygen control device, which is an important part of the integral oxygen mass transfer models.

The free-piston Stirling engine (FPSE) has been widely used in aerospace owing to its advantages of high efficiency, high reliability, and self-starting ability. In this paper, a 20-kW FPSE is proposed by analyzing the requirements of space nuclear power reactor. A code was developed based on an improved simple analysis method to evaluate the performance of the proposed FPSE. The code is benchmarked with experimental data, and the maximum relative error of the output power is 17.1%. Numerical results show that the output power is 21 kW, which satisfies the design requirements. The results show that: a) reducing the pressure shell’s thickness can improve the output power significantly; b) the system efficiency increases with the wire porosity, while the growth of system efficiency decreases when the porosity is higher than 80%, and system efficiency exhibits a linear relationship with the temperatures of the cold and hot sides; c) the system efficiency increases with the compression ratio; the compression ratio increases by 16.7% while the system efficiency increases by 42%. This study can provide valuable theoretical support for the design and analysis of FPSEs for space nuclear power reactors.

Experiments of vertically upward steam-water two-phase flow have been carried out in single-side heated narrow rectangular channel with a gap of 3 mm. Flow patterns were identified and classified through visualization directly. Slug flow was only observed at 0.2 MPa but replaced by block-bubble flow at 1.0 MPa. Flow pattern maps at the pressure of 0.2 MPa and 1.0 MPa were plotted and the difference was analyzed. The experimental data has been compared with other flow pattern maps and transition criteria. The results show reasonable agreement with Hosler's, while a wide discrepancy is observed when compared with air-water two-phase experimental data. Current criteria developed based on air-water experiments poorly predict bubble-slug flow transition due to the different formation and growth of bubbles. This work is significant for researches on heat transfer, bubble dynamics and flow instability.

A dryout mechanism model for rectangular narrow channels at high pressure conditions is developed by assuming that the Kelvin–Helmholtz instability triggered the occurrence of dryout. This model combines the advantages of theoretical analysis and empirical correlation. The unknown coefficients in the theoretical derivation are supported by the experimental data. Meanwhile, the decisive restriction of the experimental conditions on the applicability of the empirical correlation is avoided. The expression of vapor phase velocity at the time of dryout is derived, and the empirical correlation of liquid film thickness is introduced. Since the CHF value obtained from the liquid film thickness should be the same as the value obtained from the Kelvin–Helmholtz critical stability under the same condition, the convergent CHF value is obtained by iteratively calculating. Comparing with the experimental data under the pressure of 6.89–13.79 MPa, the average error of the model is −15.4% with the 95% confidence interval [-20.5%, −10.4%]. And the pressure has a decisive influence on the prediction accuracy of this model. Compared with the existing dryout code, the calculation speed of this model is faster, and the calculation accuracy is improved. This model, with great portability, could be applied to different objects and working conditions by changing the expression of the vapor phase velocity when the dryout phenomenon is triggered and the calculation formula of the liquid film.

High fidelity nuclear reactor fuel assembly simulation using CFD method is an effective way for the structure design and optimization. The validated models and user practice guidelines play critical roles in achieving reliable results in CFD simulations. In this paper, the international benchmark MATiS-H is studied carefully and the best user practice guideline is achieved for the rod bundles simulation. Then a 5 × 5 rod bundles model in the advanced pressurized water reactor (PWR) is established and the detailed three-dimensional thermal-hydraulic characteristics are investigated. The influence of spacer grids and mixing vanes on the flow and hear transfer in rod bundles is revealed. As the coolant flows through the spacer grids and mixing vanes in the rod bundles, the drastic lateral flow would be induced and the pressure drop increases significantly. In addition, the heat transfer is enhanced remarkably due to the strong mixing effects. The calculation results could provide meaningful guidelines for the design of advanced PWR fuel assembly.

The flow characteristics of reactor fuel assembly always intrigue the designers and the experimentalists among the myriad phenomena that occur simultaneously in a nuclear core. In this work, the visual experimental method has been developed on the basis of refraction index matching (RIM) and particle image velocimetry (PIV) techniques to investigate the detailed flow characteristics in China fast reactor fuel subassembly. A 7-rod bundle of simulated fuel subassembly was fabricated for fine examination of flow characteristics in different subchannels. The experiments were performed at condition of Re = 6500 (axial bulk velocity 1.6 m/s) and the fluid medium was maintained at 30 °C and 1.0 bar during operation. As for results, axial and lateral flow features were observed. It is shown that the spiral wire has an inhibitory effect on axial flow and significant intensity of lateral flow mixing effect is induced by the wire. The root mean square (RMS) of lateral velocity fluctuation was acquired after data processing, which indicates the strong turbulence characteristics in different flow subchannels. Keywords: Lateral flow mixing strength, PIV measurement, Flow characteristics analysis, China fast reactor fuel subassembly

Core-wide temperature distribution in sodium-cooled fast reactor plays a key role in its decay heat removal process, however the prediction for temperature distribution is quite complex due to the conjugate heat transfer between the assembly flow and the inter-wrapper flow. Hybrid medium model has been proposed for conjugate heat transfer modeling in the core. The core is modeled with a Realistic modeled inter-wrapper flow and hybrid medium modeled assembly flow. To validate present model, simulations for a three-assembly model were performed with Realistic modeling, traditional porous medium model and hybrid medium model, respectively. The influences of Uniform/Non-Uniform power distribution among assemblies and the Peclet number within the assembly flow have been considered. Compared to traditional porous medium model, present model shows a better agreement with in Realistic modeling prediction of the temperature distribution and the radial heat transfer between the inter-wrapper flow and the assembly flow. Keywords: Porous medium model, Hybrid medium model, Conjugate heat transfer, Inter-wrapper flow, Sodium-cooled fast reactor

ECCMIX component was introduced in RELAP5/MOD3 for calculating the interfacial condensation. Compared to other existing components in RELAP5, user experience of ECCMIX component is restricted to developmental assessment applications. To evaluate the capability of the ECCMIX component, ECCS experiment was conducted which included single-phase and two-phase thermal mixing. The experiment was carried out with test sections containing a main pipe (70 mm inner diameter) and a branch pipe (21 mm inner diameter) under the atmospheric pressure. The steam mass flow in the main pipe ranged from 0 to 0.0347 kg/s, and the subcooled water mass flow in the branch pipe ranged from 0.0278 to 0.1389 kg/s. The comparison of the experimental data with the calculation results illuminated that although the ECCMIX component was more difficult to converge than Branch component, it was a more appropriate manner to simulate interfacial condensation under two-phase thermal mixing circumstance, while the two components had no differences under single-phase circumstance. Keywords: ECC, Experiment, Condensation, RELAP5, Model assessment

Three-layer back propagation network (BPN) and genetic neural network (GNN) were developed in this study to predict the flow boiling heat transfer coefficient (HTC) in conventional and small-diameter channels. The GNN has higher precision than BPN (with root mean square errors of 17.16% and 20.50%, respectively) and other correlations. The inputs include vapor quality x, mass flux G, heat flux q, diameter D and physical parameter φ, and the predicted flow boiling HTC is set as the outputs. Influences of input parameters on the flow boiling HTC are discussed based on the trained GNN: nucleate boiling promoted by a larger saturated pressure, a larger heat flux and a smaller diameter is dominant in small channels; convective boiling improved by a larger mass flux and a larger vapor quality is more significant in conventional channels. The HTC increases with pressure both in conventional and small channels. The HTC in conventional channels rises when mass flux increases but remains almost unaffected in small channels. A larger heat flux leads to the HTC growth in small channels and an increase of HTC was observed in conventional channels at a higher vapor quality. HTC increases inversely with diameter before dry out. Keywords: Back propagation network (BPN), Genetic neural network (GNN), Flow boiling heat transfer coefficient, Conventional channel, Small channel

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.

With their rising adoption and integration into smart grids, heat pumps are becoming an increasingly important source of flexible energy. Heat pump flexibility can be utilized by using controllers to remotely manage their operation while maintaining the temperature within predefined user comfort bounds. Traditional indoor temperature modelling approaches require detailed information about the deployment site, device specific parameters and monitored data, making them inapplicable for the majority of heat pump deployments. This paper proposes a novel data-driven machine learning based method HeatFlex for indoor temperature forecasting and flexibility prediction using only 3 monitored variables: indoor and outdoor temperatures and heat pump power consumption. HeatFlex enables plug-and-play flexibility prediction from heat pumps without requiring exact device and building specifications or installation of additional sensors. This paper also introduces novel flexibility metrics enabling quantitative evaluation of heat pump flexibility prediction performance. HeatFlex is based on deep learning predictive models: Long Short-Term Memory (LSTM) and Gated Recurrent Unit (GRU) recurrent neural networks. Our experimental evaluation compared these networks with traditional multivariate linear regression and SARIMAX time series forecasting model baselines. HeatFlex performance was qualitatively and quantitatively evaluated using data from three real-world heat pump deployments with different building sizes, heat pump types and specifications. Experimental results indicate that HeatFlex is effective to accurately predict over 90% of available potential flexibility.

To pursue an environmentally friendly and efficient heating solution, a novel hybrid PV/T-air dual source heat pump based on a three-fluid heat exchanger was proposed. In the heating season, the system can absorb energy from solar and air simultaneously; in the non-heating season, the PV modules can be cooled actively or passively by heat pipe. The system was designed to provide heating and electricity for a rural single-family house in Beijing, and a mathematical model was developed. The results indicated that: (1) compared to a conventional air-source heat pump, the energy consumption of this system throughout the heating season was reduced by 13.1%; (2) for electricity generation, compared to a conventional PV panel, the electrical energy output increased by 14.7%. Additionally, the proposed integrated system has the advantages of a compact structure, standardized production, easy maintenance, and no risk of freezing.

Hyperparameter selection has always been the key to machine learning. The Bayesian optimization algorithm has recently achieved great success, but it has certain constraints and limitations in selecting hyperparameters. In response to these constraints and limitations, this paper proposed the N-RReliefF algorithm, which can evaluate the importance of hyperparameters and the importance weights between hyperparameters. The N-RReliefF algorithm estimates the contribution of a single hyperparameter to the performance according to the influence degree of each hyperparameter on the performance and calculates the weight of importance between the hyperparameters according to the improved normalization formula. The N-RReliefF algorithm analyses the hyperparameter configuration and performance set generated by Bayesian optimization, and obtains the important hyperparameters in random forest algorithm and SVM algorithm. The experimental results verify the effectiveness of the N-RReliefF algorithm.

Pressurized Water Reactor (PWR) is the most mature technology for nuclear power plants and nuclear power devices. Spray systems play an essential role in the safety of a PWR. The spray characteristics of the nozzle are of crucial importance for the performance of the spray system. In the present study, a test loop with four rotatable branches was established to obtain spray characteristics data, including flow characteristic, spray cone angle (SCA), spray flux distribution and droplet size spectrum. The effect of inclined angles was taken into consideration when designing and conducting the experiment. A specific type of pressure-swirl nozzle used in China advanced PWR was employed for the study. According to experimental data, variations of pressure drop and the spray angles versus volume flow rate were obtained, and typical curves of spray flux distribution were plotted, and the change in the flux distribution was evident with the change in inclined angle. The flux distribution results were used to build 3D spray flux distribution models together with the spray cone angles tested. The droplet size spectrum was measured by laser diffraction analysis. Typical droplet size distributions were obtained by the relevant software, and the experimental Probability Density Function (PDF) of droplet size was derived and compared with several empirical distributions.

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.

Because of the superior endurance, nuclear power has been paid more and more attention. In this article, different numerical methods were adopted to analysis thermal-hydraulic characteristics of nuclear power system under moving condition. A coupling strategy and optimal scheme of numerical calculation were designed to ensure accuracy and efficiency of computing. Based on designed theoretical model and numerical method, a self-developed program, SANPR, for analysis of moving condition effects on thermal-hydraulic characteristics of nuclear power system in civilian ship was established. Validation of core physics calculation, fuel heat conduction, system thermal hydraulic was done and the result shows that the program is able to make good prediction about natural circulation behaviour of nuclear power system under moving condition. Finally, three typical moving conditions were simulated and their effects on a designed nuclear power system was summarized.

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.

In-vessel retention (IVR) is considered as an effective mitigation strategy in realms of nuclear safety. Extensive experimental investigations have been done on the natural convection characteristics of corium pools according to the IVR concept, using various simulant materials. Based on the known COPRA (Corium Pool Research Apparatus) molten salt experiments, in which the eutectic salt (50 mol% NaNO3-50 mol% KNO3) and the non-eutectic salt (20 mol% NaNO3-80 mol% KNO3) were used as simulant materials, this paper evaluates extensively the simulant effect of simulant material on natural convection behaviors inside the corium pools through numerical simulation. The Wall-Modeled LES (WMLES) method was employed for the turbulence calculation. The distributions of corium pool temperature and heat flux as well as the crust thickness from the numerical simulations were validated with experimental data. Through numerical simulations, this work provides detailed temperature fields and velocity fields inside a large internally heated corium pool, and further discusses in depth the interplay between the heat transfer characteristics and the natural convection behaviors. This work can provide references for the selection of simulant material in the future experimental research.

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.

Nitrogen inhibition is considered as a mitigation measure against chemically sensitive mixtures in industry and carbon monoxide is possibly generated during molten corium concrete interaction in a severe accident of nuclear power plants. To study the effects of the N2 and CO on the detonation of H2-air mixtures, a detonation facility of 78 mm inner diameter and 10 m length is set up. Key measured parameters includes detonation cell size, flame velocity, lean and rich detonation limits. All the measured detonation parameters are theoretically predicted by CJ theory or one-dimensional ZND model. Detailed chemical kinetics mechanism for H2-air and H2-CO-air mixtures is coupled with reactive Euler equations. Experiments and theoretically analysis has been performed mostly at 0.101 MPa and 293 K. Results shows that N2 increases the detonation cell size and narrows down the detonable range significantly. Especially when N2 concentration is more than 43%, all mixtures are unable to detonate. Moreover, when the added N2 concentration is larger than 20%, detonation velocity does not increase with hydrogen concentration for rich mixtures. The above effects of N2 is explained by the replacement of O2 with N2 and by the weak chemical reaction of N2. CO significantly decreases the cell size of lean H2-air mixtures and increased the cell size for rich H2-air mixtures. For lean H2-air mixtures, the added CO linearly decreases the lean detonation limit, thus increasing the possibility of detonation to occur in the severe accidents. Therefore, the contribution of CO must be carefully considered in the safety assessment. Results also shows that pure CO is difficult to detonate in the air, while small quantity of hydrogen can significantly enhance the rate of CO oxidation reactions.

In present work, the two-phase flow instability between rectangular parallel channels of natural circulation under static and rolling conditions was coupled studied theoretically. Models including two-phase flow instability, natural circulation system components, and the additional force were established in combination based on the homogenous model. A computational program was written in FORTRAN language which was solved by Gear multi-value method by using control volume integrating method. The program was validated with experiments, and the results matched well with the experiment data. The marginal stability boundary (MSB) maps under different parameters were obtained by using nondimensional numbers Nsub and Npch. The influence of different kinds of pressure drop, inlet subcooling temperature of heating channels, system pressure, valve resistance, venturi flowmeters resistance, structure height, rolling condition, and the interaction effect of natural circulation and two-phase flow instability between rectangular parallel channels were analyzed. The results show that with the increase in system pressure and venturi flowmeters resistance, the system stability of natural circulation is enhanced. The influence of inlet subcooling temperature is nonlinear. The increase in valve resistance leads to the instability of system. The increase in structure height does not change system stability significantly. The rise in rolling angle and period both reduce the system stability.

The main vessel plays an important role in containing the entire primary sodium for pool-type sodium-cooled fast reactors (SFRs). The Reactor Vessel Cooling System (RVCS) has great effect on cooling the main vessel. However, little attention has been given to the study on transient characteristics of RVCS in the previous SFR research. Thus, a home-made one-dimensional (1-D) code named Reactor Vessel Cooling system Analysis Code for Sodium-cooled fast reactor (VECAS) is proposed to evaluate the thermal-hydraulic characteristics for SFR. The detailed models of the developed VECAS are presented in this paper. Moreover, the developed models have been validated against an experimental study. Numerical data of the main vessel cooling circuit are compared with the measurements of the Demonstration Fast Breeder Reactor (DFBR). The simulation results are in good agreement with the experimental data. Furthermore, the validated VECAS is coupled with the Transient Thermal-Hydraulic Analysis Code for Sodium-cooled fast reactors (THACS). The transient characteristics of RVCS in China Experimental sodium-cooled Fast Reactor (CEFR) are simulated by the coupled code. Steady analysis shows that the main vessel is cooled effectively. The peak temperature appears at the top of the main vessel lower than the permissible upper temperature limit. During the transient analysis, VECAS has predicted a reverse flow in RVCS, which contributes to the core cooling. Furthermore, sensitivity analysis of the main parameter has also been performed. Therefore, it can be concluded that coupled VECAS has the ability to evaluate the thermal-hydraulic characteristics as well as the decay heat removal capacity of RVCS. The coupled code could provide references and technical supports for the design and optimization of the pool-type sodium-cooled fast reactor.

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.

The flow condensation heat transfer of refrigerant R134a inside a multiport mini-channel having a hydraulic diameter of 0.63 mm is experimentally investigated in the present study by jet impingement cooling. Experiments were performed under the operating states of refrigerant vapor quality 0–1, heat flux 10–46 kW/m2, mass flux 115–290 kg/(m2·s), and saturation pressure 0.38–0.51 MPa. The effects of these factors on condensation heat transfer are quantitatively studied. The results indicate that the local heat transfer coefficient is seldom affected by the mass flux and vapor quality of the refrigerant but is closely related with heat flux and saturation pressure. The prediction accuracies of the existing correlations are assessed, and a new formula is proposed. This study on flow condensation heat transfer in a multiport mini-channel is meaningful for the investigation and design of a two-phase mini-channel condenser.

The present study experimentally investigated the heat transfer characteristics of flow condensation in a mini channel with serrated fins. The hydraulic diameter of the aluminum channel was 1.28 mm with 11 offset rectangular channels. The vapor of refrigerant R134a was cooled by cold water with jet impingement. After the calibration of jet impingement cooling, the influences of different parameters on condensation heat transfer were identified with the experimental conditions of vapor quality from 0 to 1, heat flux from 5 kW/m2 to 52 kW/m2, mass flux from 72 kg/(m2·s) to 145 kg/(m2·s) and saturation pressure from 0.38 MPa to 0.50 MPa. Flow patterns during condensation were also observed through the visualization design of test section. The experimental results reveal that the vapor quality and mass flux have insignificant influence on the local heat transfer coefficient. The heat transfer coefficient increases with the heat flux and decreases with the saturation pressure. Annular flow is dominant during the whole condensation process. As the existing empirical correlations fail to give reasonable prediction of the present experimental results, a new one is developed with the mean relative deviation and mean absolute relative deviation of 2.3% and 12.0%, respectively. The present work can provide valuable guidance for the design of mini channel condenser for engineering application.

For Fluoride-salt-cooled High-temperature Reactors (FHRs), a pebble bed experimental facility with internal heat generation is set up, using the Dowtherm A as the simulant fluid. Dowtherm A can match the Prandtl number of FLiBe of 520–700 °C at a low temperature range of 45–105 °C. The effects of inlet temperatures, porosity and mass flow rate are investigated on the flow and convective heat transfer characteristics. A criterion is determined for the flow regime transition from the transitional to turbulent flow. The existing correlations are unable to predict the Blake-type friction factor in the transitional flow and show mixed performance in the turbulent flow. The present empirical correlations generally underestimate the convective heat transfer coefficients of the high-Prandtl-number fluid in the pebble bed. New correlations are proposed for predicting the flow and convective heat transfer coefficients with Prandtl number at 14–19 and Reynolds number at 90–2500.

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

In this paper, the environmental control system of aircraft driven by a power turbine is further analyzed. Through the numerical simulation of the change of the bleed state under different flight conditions and the change of the flow field under different nozzle opening, the simulation results are verified by the experimental results, and the specific change rules of the power turbine output torque, power, and bleed flow are obtained. It is analyzed quantitatively that adjusting the adjustable nozzle ring can keep the output power stable, widen the flight envelope, and improve the stability of the environmental control system.

The evaluation of thermal-hydraulics of core under decay heat removal conditions is essential to the safety evaluation of a sodium-cooled fast reactor. The inter-wrapper flow (IWF) has an influence on the thermal-hydraulics of core. However, the study of the flow and heat transfer of IWF was limited. In this paper, a 2D layered IWF model was developed, and some tests were simulated to validate the IWF model, including heat removal tests SHRT-17 and SHRT-45R conducted in the experimental fast reactor EBR-II and a natural circulation test performed during the PHENIX end-of-life experiment. In order to simulate all the components of the primary system, the IWF model is coupled with the Transient Thermal-Hydraulic Analysis Code for Sodium-cooled fast reactors (THACS). In the simulation without IWF model, the predicted peak temperature of the instrumented subassembly XX10 in EBR-II is about 150 K lower than test data, and the predicted average outlet temperature of reactor core in PHENIX is about 20 K higher than test data. While the predictions of THACS with IWF model agree well with the test data. The results show that the IWF can be accurately simulated by the IWF model, and the IWF model improves the accuracy of the simulations of the reactor core. Further, some sensitivity analyses were conducted to provide better reference for the study of sodium-cooled fast reactors.

The effect of entrainment from the Hot Leg (HL) to the Pressurizer Surge Line (PZR-SL) on the characteristics of Countercurrent Flow (CCF), is experimentally studied using air-water in a transparent test section, 1/4 scaled-down model of the HL and PZR-SL assembly of AP1000. The typical CCF behaviors are recorded with video cameras for understanding of mechanisms of Countercurrent Flow Limitation (CCFL) affected by entrainment. Two entrainment flow regimes, stratified-flow entrainment and slug-flow entrainment, are found to occur under certain HL pipe water levels and air velocity conditions. The entrainment effect on the characteristics of CCF in the test section is analyzed in detail based on the two-phase air and water Kutateladze numbers, the transient hydraulic parameters and the flow behaviors. The stratified-flow entrainment does not always influence CCFL extent, unless the entrainment is strong enough and reverse flow of water occurs in the PZR-SL pipe. However, the slug-flow entrainment markedly influences CCFL extent. Two empirical correlations are developed respectively for high HL pipe water level situation with slugging entrainment and for medium and low HL pipe water level situation without slugging entrainment. The present data are compared and verified with the Small-Break Loss-of-Coolant Accidents (SB-LOCAs) tests data on AP600 Scaled Integral Test Facility (APEX). The present experimental research provide insight on the CCF flow phenomena in the HL and PZR-SL assembly for the third-generation passive nuclear reactors during the Fourth Stage Automatic Depressurization Stage (ADS-4) of SB-LOCAs and support theoretical model development and the validation of computational codes.

Tritium control is one of the most key issues for the development of Molten Salt Reactor (MSR). The tritium transport characteristics in 2MWth Thorium Molten Salt Reactor-Solid Fuel (TMSR-SF) proposed by Shanghai INstitute of Applied Physics (SINAP) is studied in this paper. Three works are conducted: Firstly, an integrated numerical model is established, including tritium production, adsorption, diffusion, corrosion and permeation. Secondly, the Tritium trAnsPort chAracteristics analysiS code (TAPAS) for TMSR-SF is developed and benchmarked. The results prove the fidelity and accuracy of TAPAS. Finally, the distributions of the key variables including TF, T2 and Cr2+ in TMSR-SF system at both steady and transient state are obtained. In addition, some methods for tritium mitigation are evaluated by TAPAS. Numerical results show that tritium can be efficiently removed from the molten salt via graphite absorption. The amount of T2 at steady state is two orders of magnitude more than TF. Tritium permeation can be prevented by increasing 7Li enrichment and reducing redox potential. Tungsten or similar low-permeability material is a candidate for tritium barrier when covered on the surface of the heat exchanger wall. This work provides some valuable considerations for the tritium control in the MSR.

HPR1000 (Hua-long Pressurized Reactor) Nuclear Power Plant is a type of third generation advanced PWR developed in China. In this study, a home-made multi-compartment containment code, ATHROC (Analysis of Thermal Hydraulic Response Of Containment), was developed to analyze the thermal-hydraulics and hydrogen behaviors in HPR1000 containment. Two types of bulk fluids, the atmosphere bulk fluid and the pool bulk fluid, were used to model the transient behaviors of multiphase flow. The atmosphere bulk fluid consists of steam, noncondensables and homogeneously dispersed liquid droplets. The code contains comprehensive models, including flow model, heat and mass transfer model, engineering safety feature model, etc. A plume model was implemented to assess buoyancy driven plume movement inside a compartment. A counter-current flow model was developed to evaluate the buoyancy driven bidirectional exchange flow through the junctions. Also, a film tracking model was developed to simulate film flow on the contiguous walls. Special methods such as using variable time steps and non-uniform nodalization for heat sinks were applied to speed up the computation. Code assessments were carried out by simulating several separated effects tests (Phebus FPT0 test, JAERI tests) and integral effects tests (CVTR tests, NUPEC M-7-1 test). Comparing results between code simulations and experimental data showed that the code was capable of providing reasonable predictions.

Sodium-cooled fast reactor (SFR) is one of the most promising reactors among the six Gen-IV reactor systems due to its significant advantages in close fuel cycle, comprehensive technology foundations and operation experiences. China is designing and constructing a demonstration SFR, in which, to ensure the reactor passive safety, a direct reactor auxiliary cooling system (DRACS) with the inter-wrapper flow is proposed as the decay heat removal system. Xi'an Jiaotong University is in charge of the DRACS analysis code development. The physical models in the DRACS are extracted and the numerical models for each component are established. The Gear method and the SIMPLE method are adopted as the primary solution algorithm. The code is developed and validated by EBR-II and PHENIX benchmarks, the results of which indicate that the code can predict experimental results very well.

Since fuel elements will be exposed to high temperature and strong irradiation for a long time, the safety of space reactor is challenged not only when an accident happens but also during the normal operation. With the full consideration of plastic deformation and high-temperature creep, a new mechanical model for fuel elements in space reactor was developed. By calculation result comparison of the new model and ABAQUS, the correctness and reliability of the new model was validated. This model was implemented in FROBA code to simulate the thermal-mechanical performance of a space reactor fuel during the long-term normal operation. The result indicated that the fuel element was safe enough during the most time of normal operation, but it experienced some periods with over high fuel temperature or low thermoelectric conversion efficiency. In addition, it was found that high temperature creep of the emitter can release the pellet-emitter contact pressure and protect the emitter from mechanical failure when power remained unchanged or changed slightly. However, the stress of emitter might exceed its strength limit during the power ramp.

The helium cooled solid breeder blanket module was proposed and designed to achieve the goal of self-sustainable tritium breeding in China Fusion Engineering Test Reactor (CFETR), which was regarded as the transition from ITER to DEMO fusion reactors. A series of preliminary analyses, including thermal hydraulics, neutronics, structural mechanics and materials, have been performed for the blanket module design. However, most literatures focus on the thermal hydraulic behaviors of high pressure helium coolant flowing in cooling channels, while the thermal hydraulic characteristics of low pressure purge gas flowing slowly in pebble beds of tritium breeding zones has not been investigated in detail. In this work, the three dimensional model of typical helium cooled solid breeder blanket module was established employing Computational Fluid Dynamics (CFD) method. The corrected Ergun equation was adopted to simulate the coolant flow features in the porous medium. The temperature and flow field distributions in the First Wall (FW), neutron multiplier zones and tritium breeding zones were achieved. The most critical parts in the blanket module during normal operation condition were illustrated. Besides, the influences of purge gas inlet velocity and tritium breeding pebble-bed ball diameter on the whole blanket module thermal hydraulic features were revealed. This work could contribute to the design and optimization of helium cooled solid breeder blanket in CFETR.

Sodium cooled fast reactor (SFR) exhibit some unique thermo-hydraulic characteristics in comparison to thermal reactors, considering that SFR have a higher core power density, special configuration of assemblies, wire-wrapped spacers, special thermo-physical properties of sodium and so on. Some design limits related to fuel, cladding and coolant temperature must be met in both the normal and transient condition. Thus, it is of great importance to obtain accurate thermal-hydraulic performance for the design and safety assessment of the SFR reactor. Therefore, a subchannel analysis code SUBAC specially for SFR wire-wrapped assemblies is developed in the present study. Sensitivity analysis of different model combinations is carried out by using ORNL 19-pin benchmark test. The calculated results show that the normalized temperature at the exit of the rod bundle calculated by the Cheng-Todreas correlation combined with Cheng-Tak turbulent mixing model is consistent with the experimental data. Subsequently, the 19-pin and 37-pin bundle benchmark problems are calculated to validate the accuracy and applicability of the code. All the calculated results agree well with the experimental data, indicating that the code is competent for steady-state calculations. Finally, SUBAC code is used for performing transient analyses of EBR-II SHRT-17 XX09 subassembly. A good agreement between calculated and experimental values of the core top sodium temperature is achieved, but slightly underestimated. In conclusion, the SUBAS program is capable of being used for subchannel analysis of SFR and some suggestions for future research are given.

Different from the square-arrayed fuel rod bundles in pressurized water reactors, the fuel subassembly for sodium-cooled fast reactors features with wire spacers helically wrapped on the fuel rod. This intricate geometry consequently imposes a significant challenge in numerical simulation. In the present study, a new marked and separated meshing method is developed to reduce the required mesh and modeling time. Instead of directly generating the complex geometry and mesh, the hexahedral wire wrapper mesh was dealt with locus equations. First, the numerical modeling method was validated by combining the experimental data of a 19-pin wire-wrapped rod bundle and the simplified Cheng and Todreas correlation. Furthermore, a comparative study of different meshing methods was conducted. Then, an investigation on the three-dimensional flow phenomenon and heat transfer characteristics of a 61-pin wire-wrapped fuel subassembly for the China Experimental Fast Reactor was carried out. Finally, complex axial and transverse flow behaviors and temperature field in different types of subchannels were quantitatively analyzed. Results indicate that the axial velocity exhibits periodic oscillation with a maximum normalized value of 1.54. The flow oscillation in the interior subchannels presents a smaller amplitude and higher frequency than that in the corner and edge subchannels. Local hot spots occur in the gaps beneath the spacer wires. Sweeping and cross flow patterns exist in the peripheral and interior subchannels, respectively. In addition, the secondary transverse flow induced by the spacer wires is stronger in the corner subchannel. The cross flow strength reaches a peak value equal to 30% of the average axial velocity, which promotes the thermal mixing.

Flow blockage accidents in rectangular fuel assembly were investigated by three-dimensional CFD method in detail. Totally six coolant channels of the fuel assembly were modeled. Meanwhile, in order to make understand of the thermal hydraulic characteristics of the flow blockage accidents comprehensively, three coolant channels which included the obstructed channel were investigated and analyzed thoroughly based on simulation results. RELAP5 calculation was performed here to compare with CFD simulation results under non-blocked condition, and the comparison results indicated approximate agreement of such two types of results. On the basis of the CFD simulation, velocity and temperature profiles were discussed for some typical blockage cases, and conclusion was drawn that the redistribution of the mass flow rates occurred after the formation of the blockage, and due to the formation of obstruction, temperature of the coolant and the fuel increased rapidly which caused higher peak temperature in the blockage channel. Simultaneously, the increasing flow resistance would lead to the existence of jet-flow and reverse flow in the obstructed channel. In addition, DNBR calculation indicated that heat flux on the cladding surface would not exceed critical heat flux, thus DNB would not occur under the investigated situation.

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.

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.

After the Fukushima Dai-ichi nuclear power plant severe accident, the advanced facilities that could improve the nuclear safety performance are preferred by the public. This paper discusses the preliminary design and thermal-hydraulic analysis of a heat pipe cooled passive heat removal system for the transportable fluoride-salt-cooled high-temperature reactor of 20 MWth proposed by Massachusetts Institute of Technology in 2014. Under severe accident, the high temperature fuel and fluoride salt sustain at the reactor core, the heat pipes are inserted by gravity to remove the decay heat to the final heat sink of silo cooling system. To illustrate the feasibility of the designed passive heat removal system, the heat pipe startup performance is firstly numerically investigated by self-developed code HPTAC (Heat Pipe Transient Analysis Code). Four heat transfer limits are adopted as criteria for the success of heat pipe operation, 1) Viscous 2) Entrainment 3) Sonic 4) Capillary limits. The benchmark of heat pipe startup compared with experimental data is conducted as well as a sensitivity study for heat pipe inclination angle. Secondly, the transient performance of passive heat removal system based on reasonable assumptions of heat pipe model is analyzed. Results show that the heat pipe reaches the normal operation in 6 min, in good agreement with experimental data with the maximum discrepancy of 16%. Most of fuel decay heat can be removed in 5 h by heat pipe passive heat removal system and all the key temperature are kept below the allowed temperature limits.