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To address the efficiency bottleneck encountered in reactor design calculations for the newly developed lead-based reactor neutronics analysis code system MOSASAUR, we recently developed acceleration functions based on various coarse-mesh finite difference (CMFD) methods and spatial domain decomposition parallel algorithm. However, the applicability of these improvements to different lead-based reactors remains to be analyzed. This work collected and established core models for various types of lead-based reactor. Based on different SN nodal transport solvers, we analyzed the acceleration performance of different CMFD methods, different CPU cores, and the combination of CMFD and parallel calculation. The results indicated that the impact of different CMFD acceleration or parallel acceleration on the calculation accuracy was negligible; the rCMFD method had good stability and convergence rate, achieving speedup of several to dozens; parallel efficiency was related to the number of meshes, and for large reactor cores, superlinear speedup was achieved with 200 CPU cores; rCMFD and parallel computing could achieve combined speedup, with 200 cores achieving speedup of hundreds to thousands, typically completing a reactor core transport calculation in 1 min. [ABSTRACT FROM AUTHOR]

Lead-based reactor nuclear power has the advantages of inherent safety, compact size, light weight, long service life, and high efficiency, and it is widely applied to advanced marine equipment, unmanned underwater vehicles, and deep-sea space stations for energy and power purposes. Conducting research on lead-based reactor marine nuclear power is key to the technological innovation of energy and power for marine equipment in China. This study clarifies the application scenarios and development demands of marine nuclear power, summarizes the development status of lead-based reactor nuclear power from the aspects of technical characteristics, representative applications, and research status, and analyzes corresponding key technologies including nuclear fuel and cladding materials, key components in the primary loop, coolant processes and oxygen control, and advanced power generation technologies. Moreover, the study explores the development challenges for lead-based reactor marine nuclear power and proposes there feasible technical routes: low-temperature lead-based reactor marine nuclear power, high-temperature and high-efficiency lead-based reactor marine nuclear power, and integrated natural-circulation lead-based reactor marine nuclear power. Furthermore, we propose the following suggestions to promote the leapfrog development of marine nuclear power in China:(1) strengthening research and development (R&D) of specialized technologies for the lead-based reactor marine nuclear power and accelerating the deployment of demonstrative projects; (2) incorporating lead-based reactors into the country's major energy strategies and creating a standards system; (3) establishing national joint R&D centers and creating new models for technological innovation and industrial development.

To address the efficiency bottleneck encountered in reactor design calculations for the newly developed lead-based reactor neutronics analysis code system MOSASAUR, we recently developed acceleration functions based on various coarse-mesh finite difference (CMFD) methods and spatial domain decomposition parallel algorithm. However, the applicability of these improvements to different lead-based reactors remains to be analyzed. This work collected and established core models for various types of lead-based reactor. Based on different SN nodal transport solvers, we analyzed the acceleration performance of different CMFD methods, different CPU cores, and the combination of CMFD and parallel calculation. The results indicated that the impact of different CMFD acceleration or parallel acceleration on the calculation accuracy was negligible; the rCMFD method had good stability and convergence rate, achieving speedup of several to dozens; parallel efficiency was related to the number of meshes, and for large reactor cores, superlinear speedup was achieved with 200 CPU cores; rCMFD and parallel computing could achieve combined speedup, with 200 cores achieving speedup of hundreds to thousands, typically completing a reactor core transport calculation in 1 min.

• Formation and growth of fe-cr spinel layer follow a diffusion-controlled process. • Spallation and dissolution significantly degrade the fe-cr spinel layer over time. • Complex interplay between direct dissolution and spallation of oxide layer observed. • Microstructural evolution includes localized formation of Fe 3 O 4 and role of grain structure. • The dissolution rates of steel in LBE include contributions from direct dissolution and oxide scale loss. The application of Liquid Lead-Bismuth Eutectic (LBE) as a coolant in Lead-cooled Fast Reactors (LFRs) presents significant challenges due to its corrosive nature, especially at elevated temperatures. This study investigates the corrosion kinetics and mechanisms of 15–15Ti stainless steel in flowing LBE at 500 °C with oxygen concentrations ranging from 1 to 3 × 10−6 wt% and a flow rate of 1 m s-1. The research highlights the formation and growth of the Fe-Cr spinel oxide layer, which follows a parabolic rate law indicative of a diffusion-controlled process, crucial for long-term material stability predictions. Despite the initial effectiveness of the Fe-Cr spinel layer in mitigating corrosion, its integrity is compromised over time due to spallation and dissolution, allowing penetration of LBE elements and selective dissolution of Fe, Ni, and Cr. The study reveals a complex interplay between direct dissolution of steel components and spallation of the oxide layer, providing critical insights into material loss mechanisms and the degradation of structural materials in LBE-cooled reactors. [Display omitted] [ABSTRACT FROM AUTHOR]

Corrosion tests of T91 were carried out in the stagnant lead–bismuth eutectic (LBE) containing 10−6, 10−7, and 10−8 wt% oxygen at 500°C for up to 2000 h, respectively. The dependence of corrosion characteristics on oxygen concentrations in LBE was obtained. The results show that the oxide‐scale structure changes from a three‐layer magnetite/spinel/internal oxidation zone (IOZ) scale under the oxygen concentration of 10−6 wt%, to the formation of a two‐layer spinel/IOZ scale under the oxygen concentrations of 10−7 and 10−8 wt%. In addition, with the decrease of the oxygen concentration, the Cr/Fe ratio in the Fe–Cr spinel slightly increases in a linear rule while the thicknesses of the oxide layers gradually decrease. [ABSTRACT FROM AUTHOR]

This editorial discusses the technological advancements in Gen4 nuclear energy systems and small reactors. These advanced systems, such as sodium-cooled fast reactors, lead-cooled fast reactors, and gas-cooled fast reactors, offer superior sustainability, safety, reliability, and economics compared to current-generation systems. The article highlights specific research on the safety analysis of a lead-based reactor, the operational characteristics of a micro gas-cooled reactor, the flow stability of supercritical carbon dioxide in a Brayton cycle, and the fault diagnosis of rotating machinery using deep learning models. The authors hope that these articles will serve as a valuable reference for researchers in the field of Gen4 nuclear energy systems and small reactors. [Extracted from the article]

The interactions and energy transfer between liquid metal and water are of great research value in developing and utilizing clean energy (solar and nuclear). Understanding how much energy transfer exists in interactions and the consequences of these interactions is crucial for ensuring the stability of energy development and response measures in case of accidents, especially in third-generation ultra-high temperature concentrated solar power and fourth-generation nuclear energy systems. In this paper, a jet visualization test was performed to investigate the interaction characteristics of liquid metal with water. We found that the energy transfer between liquid metals and water is constrained, distinguishing three distinct pressurization modes due to mass transfer, while also developing an energy transfer model for liquid metals. First, the visualization data show that liquid metal-water interaction consists of four stages: coarse mixing, large steam bubble expansion and fragmentation, direct contact and bubble migration. This process is accompanied by the splashing of metal droplets and shaking of the free liquid surface. Furthermore, the water escapes from the melt pool under some conditions, and modal boiling occurs at the surface of the melt pool, hindering heat exchange. Subsequently, three cover gas pressurization modes were distinguished according to the different interactions of liquid metal with water. Finally, based on the perspective of restricted energy transfer, the liquid metal temperature drop model was established, and there is an appropriate matching between the model and test data with ±20 % uncertainty. • We discovered the restricted energy transfer between liquid metal and water. • Based on the perspective of restricted energy transfer, the liquid metal temperature drop model was developed, which matches the test data better. • The phase behavior and sudden increase characteristics in pressure in the interaction between liquid metals and water were clarified. [ABSTRACT FROM AUTHOR]

The energy conversion and utilization of liquid metals is becoming one of the effective ways to solve energy problems. The conversion and transfer of thermal energy from liquid metals to water are significant in advanced nuclear energy and metallurgy. It has been an urgent concern, especially in Generation IV of advanced nuclear energy systems‑lead-cooled fast reactors and the ultimate energy source for humanity-fusion engineering. In this paper, a large-scale liquid metal and water interaction test platform was developed to investigate the lead‑bismuth liquid metal and water interaction and the cover gas pressurization characteristics. Firstly, the evolution of phase behavior and water/vapor migration paths were inferred based on the transient temperature law. Secondly, the cover gas pressurization patterns above the molten pool were analyzed, and the test data showed three pressurization patterns, including two types of local vapor explosions. Additionally, this paper predicted the existence of another dangerous pressurization mode, i.e., overall vapor explosion. Finally, the cover gas pressurization rate model was established, and there is an appropriate matching between the model and test data with an average relative error of 12.6%. This paper provides an essential reference for profoundly understanding the energy conversion and thermal interactions between liquid metals and water. • A large-scale liquid metal and water interaction test platform was developed to investigate energy conversion and interaction. • Based on the transient temperature law, we discovered the gas-phase evolution characteristics during lead-water interaction. • Three different types of pressurization patterns of cover gas were identified. • A model of cover gas pressurization was developed to predict the pressurization rate. [ABSTRACT FROM AUTHOR]

Summary: A reheating process is generally incorporated in a supercritical CO2 (S‐CO2) Brayton cycle to enhance its efficiency. The heat transfer process from the reactor coolant to the working fluid of the power cycle is a key issue encountered when designing reheating power systems for the lead‐based reactor. The traditional reheating system, called RH‐1, utilizes an intermediate coolant circuit. In this paper, a novel reheating system, called RH‐2, is proposed. It eliminates the intermediate coolant circuit and combines the processes of the primary heating and reheating in a single heat exchanger. A thermodynamic analysis of three different systems for the lead‐based reactor integrated with the S‐CO2 power cycle with or without reheating was conducted to evaluate the performance of the proposed system. The results confirmed that the performance of RH‐2 was the best of all the three systems. Under the same reactor conditions, the system efficiency of RH‐2 was greater than those of RH‐1 and the recompression (no reheating) system by 1.2% and 1.7%, respectively. RH‐2 could also maintain higher efficiency when the main operating parameters varied. The efficiency of RH‐2 was higher at different core outlet temperatures and split ratios. The maximum efficiency at optimal maximum pressure of RH‐2 was greater than those of the other two systems. RH‐2 was less sensitive to the variations in the isentropic efficiencies of the components than the other two systems, while the turbine isentropic efficiency demonstrated a significantly higher impact on the system efficiency than the two compressors (approximately 3.8 times). [ABSTRACT FROM AUTHOR]

The China lead-based research reactor (CLEAR-I) with 10 MW(thermal) will be built for the research of the accelerator-driven subcritical system (ADS) and lead-cooled fast reactor technology. Compared with other reactors, ADS has a spallation target placed in the reactor core center, which affects the refueling motion in the vessel. It is difficult to reach the fuel assemblies near the proton beam tube through the common vertical refueling gripper. In this technical note, a new cantilever-type and internal grasping gripper with defined refueling process was designed to handle all assemblies in CLEAR-I. The static and kinematics analyses of the gripper were carried out by ANSYS Workbench and MATLAB considering the influence of working in air and liquid metal. The structural and kinematic simulation results show that the design of the gripper is feasible for CLEAR-I. It provides an advanced refueling solution for the technical validation of ADS. [ABSTRACT FROM AUTHOR]

Based on the pool type lead-based reactor, the two-phase flow phenomenon remains a longstanding challenge in a Steam Generator Tube Rupture (SGTR) scenario. In order to investigate the bubble rising velocity in opaque liquid Lead Bismuth Eutectic (LBE), experiments were carried out by injecting argon gas into four transparent liquids (water, glycerol, alcohol and FC-3283). The appropriate drag coefficient of a single bubble was obtained by dimensionless Eötvos number under different liquids. For bubble equivalent diameter from 4 mm to 6 mm, the bubble flow located in the regime of surface tension dominant, where the bubble terminal rising velocity tended to a value almost independent of the bubble diameter. The single bubble rising velocity remained about 0.252 m/s to 0.274 m/s in gas-liquid metal bubbly flow. An appropriate modified correlation was recommended to roughly forecast the single bubble terminal rising velocity in two phase flow with high liquid-to-gas density ratio. These data can provide a reference on analyzing bubble migration process under accident condition. [ABSTRACT FROM AUTHOR]

The temperature fluctuations induced by incomplete mixing of coolants with different temperature may cause thermal fatigue at the components of the lead-based reactor core outlet. Thus the accurate analysis of the phenomenon is very crucial for reactor safety operation. In this paper, the temperature fluctuations of the lead-based reactor core outlet were simulated by using large eddy simulation (LES) method in the simplified core outlet models. In order to analyze the temperature fluctuation sensitivity for the fuel assembly design parameters, such as the fuel assembly size and the gap between two adjacent fuel assemblies, five geometry models were constructed with different fuel assembly design parameters. The time histories of temperature fluctuations at different monitoring points on the center of three fuel assemblies were obtained. Then the amplitudes and the power spectrum density (PSD) of temperature fluctuations were analyzed, in order to compare temperature fluctuations of different geometry models at the same locations of core outlet. Finally the distribution characteristics of core outlet temperature fluctuations were obtained in axial directions, and the temperature fluctuation sensitivity with fuel assembly parameters was also analyzed based on the amplitudes, PSD and the normalized root-mean square temperature analysis. It is found that the temperature fluctuation intensity is enhanced with the increase of the gap size between adjacent fuel assemblies and the opposite edge width of each fuel assembly. The analysis results could provide important references for optimized design and engineering guidance of lead-based reactor. [ABSTRACT FROM AUTHOR]

It is of great significance to study the thermal-hydraulic characteristics of Lead-based Fast Reactor. In these years, the computational fluid dynamics method has become an attractive way in thermal-hydraulic and safety simulations of liquid metal cooled reactors. In order to investigate thermal-hydraulic characteristics of single-phase natural circulation flow in liquid metal cooled reactors, a self-developed CFD code was developed in this paper. The new code was proposed based on the two-phase compressible model, including liquid and vapor governing equations and a series of auxiliary models. The commercial CFD software ANSYS FLUENT was used to verify and validate the new code by comparing their simulation results under a typical heat driven cavity flow case. The comparison showed that the results simulated by the new code had good agreement with the ones calculated by ANSYS FLUENT. Furthermore, by coupling the above CFD model with the fuel pin heat transfer model, the application was conducted based on a full-scale reactor steady-state simulation of a natural circulation driven LBE-cooled fast reactor. From the results, it can be concluded that the new code can simulate and evaluate thermal-hydraulic behaviors of the LBE-cooled reactor well. [ABSTRACT FROM AUTHOR]

A method is given to predict the steam penetration depth under an incidental Steam Generator Tube Rupture (SGTR) accident. Several similar experiments were performed by injecting gas into water to simulate the steam jetting into liquid Lead Bismuth Eutectic (LBE). The steam penetration behaviors including flow regimes and cavity phenomena were captured by visualization method and a semi-empirical correlation was modified for the penetration depth based on dimensional analysis and experimental data. The results showed that the penetration depth was proportional to the density ratio of gas jet to liquid coolant, gas velocity, and break size of tube. Further predictions with similarity theory have been conducted. The penetration depth was about 0.7 m during a large break (a double-ended break) of one SG cooling tubes under a postulated SGTR accident of lead-based reactor. [ABSTRACT FROM AUTHOR]

To avoid the damage of the fuel rod cladding when a fuel assembly (FA) is totally blocked, a special configuration of the fuel assembly spike was designed with some equalizing holes in the center region which can let the coolant to flow during the totally blockage scenarios of FA. To study the hydrodynamic characteristics between the equalizing holes and an appropriate resistance, a RELAP5 model was developed for a 10 MWth lead-based reactor which used lead-bismuth as coolant. Several FA total blockage and partial core blockage scenarios were selected. The simulation results indicated that when all the FA spike equalizing holes had the same hydraulic resistance, only a narrow range of suitable equalizing holes resistances could be chosen when a FA was blocked. However, in the two or more FA blockage scenarios, there were no appropriate resistances to meet the requirement. In addition, with different FA spike equalizing holes with different resistances, a large range of suitable equalizing hole resistances could be chosen. Especially a series of suitable resistances were selected when the small power FA resistance was 1/2, 1/4, 1/8 of the large one. Under these circumstances, one, two or three FA blockages would not damage the core. These demonstrated that selecting a series of suitable hydraulic resistances for the equalizing holes could improve the safety characteristics of the reactor effectively. [ABSTRACT FROM AUTHOR]

Thermal striping is one of great safety concerns for liquid metal cooled fast reactor, and its detailed analysis is thus very important for reactor design and operation. In this paper, large eddy simulation (LES) approach was employed for the accurate simulation of temperature fluctuations for thermal striping in double jet model by using FLUENT. The normalized amplitudes of temperature fluctuations of lead and Lead–Bismuth Eutectic (LBE) were obtained. Then in order to compare temperature fluctuations of lead-based reactor and that of sodium cooled reactor, the same calculation of sodium was also performed. Moreover, the power spectrum density (PSD) of temperature fluctuation of lead, LBE and sodium was analyzed by using MATLAB for investigating the dominant frequencies of temperature fluctuation of the three fluids. It is concluded that both amplitudes and frequencies of lead-based materials fluctuations are larger compared with that of sodium temperature fluctuations, and thermal striping should thus be paid more attention to in lead-based reactor than sodium cooled reactor. [ABSTRACT FROM AUTHOR]

China Lead-based Research Reactor (CLEAR-I) will be built to develop the technologies of lead–bismuth eutectic cooled reactor. The In-Vessel Fuel Handling System (IVFHS) can achieve all positional refueling of the core by the complex operation of the double rotating plugs, the transfer machine and the rotor lift machine. One five degrees of freedom (5-DOF) rod mechanism of the IVFHS was proposed. Each joint parameter of the kinematic pairs was decided by the D-H method. The kinematics equation and velocity Jacobi matrix of the 5-DOF rod mechanism were deduced by the kinetic analysis. The rod position, rotation angle and velocity can be deduced by the moving trajectory of the transfer machine gripper. The feasibility was verified by the computer simulation and the principle prototype. [ABSTRACT FROM AUTHOR]

The multi-modular reactor, which connects multiple reactor units to drive a single turbine, is considered to be an innovation in the nuclear power system. The load factors of reactor units need to be adjusted in the event of load changes and module scram, which can cause fluctuations in numerous parameters and affect the stability of the reactor. This paper discusses the effective matching operation problem between reactors. To realize an orderly adjustment of load factors, an automatic power controller incorporating two operational strategies "extreme first" and "equal change" is designed. The feedforward-feedback control method is introduced to regulate the average temperature of the secondary circuit in order to reduce the control delay error caused by the thermal inertia of the system. An operation and control simulation of the multi-modular lead-based reactor based on the China lead-based research reactor (CLEAR-I) is performed. The numerical results show that the system parameters are well controlled, and the effective coordination of each module is realized. Image 1 • Load changes and module scram can affect the stability of the multi-modular reactor. • An automatic power controller is designed to match the operation between reactor units. • The proposed feedforward-feedback control method can reduce the control delay error. [ABSTRACT FROM AUTHOR]

A conceptual design of compact and highly efficient heat exchanger has been developed and proposed for the 10 MW th lead-based reactor using mathematical and numerical methods. Lead Bismuth Eutectic (LBE) and Pressurized Liquid Water (PLW) are selected as working fluid across the hot and cold sides of the Compact Heat Excahnger (CHE). The mathematical method based on energy balance equations has been used to develop the design of CHE with complete specifications of structural parameters. While, in the numerical method, FLUENT has been used to analyze and evaluate the detailed observation of thermal performance, Stanton number, and development of the irreversible heat transfer by the Bejan number. The thermal performance of CHE can be controlled with the flow rate of the PLW. The maximum overall heat transfer coefficient and effectiveness were 7528 W ⋅ m−2K−1 and 0.79, respectively. The thermal-hydraulic analysis involves the assessment of Nusselt number (Nu), temperature contour, and velocity vector and stream line. The numerical simulation results are agreed well with the correlation and the error is within ranges of ~ 2- 8%. This study can be used as a reference for the development and design of a compact and efficient heat exchanger for lead-based reactors. [ABSTRACT FROM AUTHOR]

• Blockage accident of lead-based reactor was simulated. • The neutronics/thermal-hydraulics coupling model was established. • The effect of axial position for blockage was analyzed. Lead-based reactor coolants have many excellent characteristics but the coolants may seriously erode the fuel cladding and the structural materials and the flow area will be reduced by the corrosion products. This phenomenon is called blockage accident. In this paper, the blockage accident of single subassembly for a lead-based reactor has been analyzed by applying the commercial computational fluid dynamics (CFD) software coupled with the point reactor kinetics model. Firstly, the neutronics/thermal-hydraulics coupling model was established by user defined function (UDF) of FLUENT software based on the point reactor kinetics function with one group delayed neutron. Then the neutronics/thermal-hydraulics coupling model was validated with a single fuel pin model by comparing the fuel power, the cladding temperature and the coolant temperature with the calculation results of the RELAP5 code. Secondly, after considering mesh independence, turbulence models and other factors, the steady state of a single subassembly with 61 fuel pins at full power for a lead-based reactor designed by INEST/FDS Team was simulated in the normal operating conditions. Finally, a single subassembly for the center blockages of the flow channel was simulated by the neutronics/thermal-hydraulics coupling model. The temperature distributions and the maximum temperature were compared with that of the normal operating conditions and the effect of axial position for blockage was analyzed. It shows that the core outlet temperature and the highest temperature of cladding are increasing after blockage accident happening. And the highest temperature of cladding is lower calculated by neutronics/thermal-hydraulics coupling model than the condition without considering reactivity feedback. And the maximum cladding temperature for the 500 mm axially location blockage accident is higher. [ABSTRACT FROM AUTHOR]

Natural circulation and forced circulation are two types of normal operation conditions that usually appeared in nuclear power systems. However, it should be noted that the reactor safety and reliability can be jeopardized due to significant fluctuation of the mass flow rate through the lead-based reactor core during the conversion from forced to natural circulation. Therefore, a model of 10MW th lead-based reactor based on RELAP5 mod4.0 is developed to analyze the effects of the five factors on the transient characteristics, including the power control system, the initial steady-state power level, the human intervention, shutting down each main pump one by one and the bypass of the main pump separately. The results show that using power control system can obviously reduce the transition time and realize the rapid response of the reactor during the conversion. Moreover, a higher initial power level results in a larger power fluctuation, a larger maximum value of cladding peak temperature and a stronger natural circulation. As a result, the maximum initial power level can be obtained according to the cladding temperature limitation (823 K). Especially, the optimized control strategy taking into account the combined effects of three factors, namely human intervention, shutting down each main pump one by one as well as the bypass of the main pump, can increase the maximum initial power level from 45%FP to 75%FP. • A RELAP5 model for a 10MWth lead-based reactor (CLEAR-I) was built to study the transient characteristics from forced circulation to natural circulation. • The effects of the five factors on the transient characteristics were analyzed separately. • The optimized control strategy taking into account the combined effects of three factors can increase the maximum initial power level from 45%FP to 75%FP. [ABSTRACT FROM AUTHOR]