Lead-bismuth Fast Reactor Research Articles

In-situ measurement of hydrogen isotope behavior in high temperature LBE: Diffusivity, permeability, and solubility

The Lead-Bismuth Fast Reactor (LBR) emerges as a promising concept due to its superior neutron economy, chemical stability, and thermal properties. From the nuclear safety standpoint, the focus has predominantly been on the behavior of 210Po and other fission products, while the issue of tritium in LBRs has not been sufficiently addressed due to the inconvenience of tritium experiments. Formation of tritium/helium bubbles induces significant local stress and volume expansion, leading to hardening and embrittlement of structural materials, thus expediting their degradation through irradiation effects. Current understanding of tritium transport within liquid Lead-Bismuth Eutectic (LBE) remains incomplete. To bridge this gap, a novel device employing the "permeation pot" method has been developed for the first experimental quantification of diffusivity, permeability, and solubility of hydrogen isotopes in liquid LBE. Notably, hydrogen diffusivity in this medium is found to be three to four orders of magnitude greater than in conventional 316L stainless steel structural material. Furthermore, the temperature-dependence of diffusivity in liquid metals is minimal compared to solids, as indicated by the activation energy. Conversely, solubility in 316L significantly surpasses that in LBE by three to four orders of magnitude. This discrepancy accelerates the tritium release from LBE to structural material, leading to the failure of the structural material.

Steady and transient analysis on thermal hydraulic characteristic of helical coiled once-through steam generator of liquid metal fast reactor

A transient model for thermal–hydraulic characteristics of helical coiled once-through steam generator (HCOTSG) of liquid metal fast reactor was established based on global discrete grid method. The model meshed the liquid metal circuit and water-steam circuit. Meanwhile, unsteady heat conduction equations were built to accurately simulate coupled heat transfer between two sides. Four-equation drift-flux method was adopt for water-steam flow. Taking lead–bismuth fast reactor as example, thermal–hydraulic characteristics of HCOTSG were first calculated under steady conditions. It was found that heat flux distribution along steam generator is extremely uneven, and the difference between maximum and minimum wall heat flux reaches hundreds of times. Then, the transient response was simulated when inlet conditions of primary side were perturbed by step changes, and it was found that maximum wall heat flux increases from 1175.5 kW/m2 to 1365 kW/m2, increasing by nearly 16 %, when inlet lead–bismuth temperature step increases from 450°C to 480 °C.

Numerical study on heat transfer characteristics of LBE cross flow tube bundle under rolling conditions

Combined with the neutron economy and good heat transfer performance, lead–bismuth fast reactor (LFR) has a larger prospect for marine applications. Helical coiled once-through tube steam generator (HCOTSG) is also a typical heat exchanger option for LFR. It is widely adopted in a variety of energy utilization fields, such as petrochemical industry. Study on Lead-bismuth eutectic (LBE) cross flow of tube bundle under moving conditions is necessary for obtaining the flow and heat transfer characteristics of LBE HCOTSG in marine conditions. A numerical method to simulate the heat transfer characteristic of LBE cross tube bundle flow under rolling conditions is proposed. The influence of the rolling conditions on the models with different tube arrangements are investigated. The temperature fluctuation curves with time and the time-averaged heat transfer characteristics are compared and analyzed. The results of the influence of rolling conditions under different working conditions and different tube bundle arrangements are obtained. The improvement ratio under inclined arrangement and staggered arrangement is more than 20 %, but the improvement under inline arrangement can only reach 10–15 %. Among the calculated models under different rolling conditions, the maximum heat transfer improvement reached 27.3 %. Besides, the circumferential temperature results under different conditions were also compared to analyze the influence of rolling conditions. This study may contribute to the application of LBE HCOTSG under marine conditions.

Mesh morphing-based pre-processing and numerical simulation of blockage accident in lead–bismuth fast reactor fuel assembly

This paper presents a CFD modeling approach for a 19-rod bundle fuel assembly with spiral semi-cylindrical ribs, which potentially can be used for liquid metal cooled fast reactor, as well as a numerical simulation of flow and thermal behavior under partially blocked conditions. The ribs on fuel rod surface are integrally formed with the cladding to enhance heat transfer while preventing detachment. However, the complex channel geometry presents challenges for grid generation, as using a large number of tetrahedral or polyhedral elements would consume significant computational resources. Therefore, this paper proposes a mesh morphing-based pre-processing method to establish a structured hexahedral mesh for such complex geometry. Using this approach, the 19-rods assembly is analyzed, and the result has been verified by conventional mesh scheme as well as experiment.

Design and analysis on the HP-PHRS for small modular lead-bismuth fast reactor

The heat pipe is an efficient heat-conducting element that can efficiently and passively transfer heat from a high-temperature heat source to a low-temperature heat source. It is widely used in the various industries. Based on high-temperature heat pipe technology, this paper designs a passive decay heat removal system for a small modular lead–bismuth fast reactor, and uses the system code ATHLET coupled with the heat pipe calculation module to calculate and analyze the thermal–hydraulic characteristics of the reactor under station black out accident.

Optimization study of lead-based reactor heat exchangers based on Dakota coupling CFD

The lead-bismuth fast reactor is one of the fourth-generation reactor types.The microchannel heat exchangers are used as intermediate heat exchangers, which significantly influence the efficiency of supercritical carbon dioxide (S–CO2) Brayton cycle power generation. It is necessary to quantify the uncertainties and optimize the design of heat transfer and pressure drop of the microchannel heat exchangers. Due to the high cost and time-consuming problem of numerical simulation, this study employs the Arbitrary polynomial chaos expansion method to generate surrogate models for simulation calculations. Specifically, the microchannel parameters of the heat exchanger, namely flow direction spacing, transverse spacing, staggered spacing, and inlet velocity, are studied for the simulation. The pressure drop and convective heat transfer coefficient of the heat exchanger are considered as the objective functions for both simulation and optimization models. Uncertainty quantification analysis, sensitivity analysis, and optimization design of those parameters are conducted. By coupling DAKOTA with CFD, an optimization objective function is established to obtain optimal design data within specified ranges.

Development and validation of the lead-bismuth cooled reactor system code based on a fully implicit homogeneous flow model

The liquid lead-bismuth cooled fast reactor has been in a single-phase, low-pressure, and high-temperature state for a long time during operation. Considering the requirement of calculation efficiency for long-term transient accident calculation, based on a homogeneous hydrodynamic model, one-dimensional heat conduction model, coolant flow and heat transfer model, neutron kinetics model, coolant and material properties model, this study used the fully implicit difference scheme algorithm of the convection-diffusion term to solve the basic conservation equation, to develop the transient analysis program NUSOL-LMR 2.0 for the lead-bismuth fast reactor system. The steady-state and typical design basis accidents (including reactivity introduction, loss of flow caused by main pump idling, excessive cooling, and plant power outage accidents) for the ABR have been analyzed. The results are compared with the international system analysis software ATHENA. The results indicate that the developed program can stably, accurately, and efficiently predict the transient accident response and safety characteristics of the lead-bismuth fast reactor system.

Core design and neutronic analysis of a long-life LBE-cooled fast reactor NCLFR-Oil

The exploitation of heavy oil reservoirs calls for the implementation of thermal recovery techniques. To provide the heat source for oil recovery energy system, NCLFR-Oil, a Natural Circulation Lead-bismuth Fast Reactor was designed for heavy oil extraction, with a designed power of 40 MW and a lifetime of at least 15 years. In order to achieve the designed object for a reactor core that is durable, compact, transportable, and exhibits strong passive and active safety features, the selection of core materials such as fuel and absorber, evaluation of reactivity control system and corresponding neutronic performance have been optimized by iterative design in this paper with the reference of many small modular reactors. The depletion calculation confirms that the designed reactor can operate for 19 years without refueling. Neutronic performance were analyzed for multiplication factor within lifetime, isotopes conversion, neutron energy spectrum, radial and axial power distributions, nuclear fission reaction rates, reactivity feedback coefficients, and control rod worth with shutdown margins. The core analysis shows that the reactor has good power and safety characteristics, which can fully meet the energy demand of heavy oil exploitation.

Heat transfer evaluation of liquid lead-bismuth eutectic cross flow tube bundle: Experimental part

Helical coil once-through tube steam generator (HCOTSG) is an excellent form of heat exchanger. HCOTSG is adopted in various industries, such as Lead-bismuth Fast Reactor (LFR). The flow and heat transfer in the shell side of HCOTSG are important for its design and optimization. It is of great significance to carry out relevant experiments to obtain empirical correlations. In this paper, experimental study on the lead-bismuth eutectic (LBE) cross flow around the tube bundle is carried out to investigate the thermal-hydraulic characteristics based on geometric approximation of local region. Three test sections with different inclination angles of 0°, 4°, 8° are adopted in the experiments with Peclect number (Pe) up to 350. The experimental sections are uniformly heated with heat flux ranging from 20 to 100 kW/m2. The heat transfer characteristic of LBE cross flow around inclined tube bundle is obtained, considering the effects of inclination angle of the test section. The heat transfer correlation of LBE cross flow around inclined tube bundle is proposed for the first time. By comparing the data of the experiments, the results are all within the error range of 10%. The correlation is appropriate for the program development of LBE HCOTSG. This study can provide essential experimental data for the design and code development of LBE HCOTSG.

Development of reduced-order thermal stratification model for upper plenum of a lead–bismuth fast reactor based on CFD

After an emergency shutdown of a lead-bismuth fast reactor, thermal stratification occurs in the upper Plenum, which negatively impacts the integrity of the reactor structure and the residual heat removal capacity of natural circulation flow. The research on thermal stratification of reactors has mainly been conducted using an experimental method, a system program, and computational fluid dynamics (CFD). However, the equipment required for the experimental method is expensive, accuracy of the system program is unpredictable, and resources and time required for the CFD approach are extensive. To overcome the defects of thermal stratification analysis, a high-precision full-order thermal stratification model based on CFD technology is prepared in this study. Furthermore, a reduced-order model has been developed by combining proper orthogonal decomposition (POD) with Galerkin projection. A comparative analysis of thermal stratification with the proposed full-order model reveals that the reduced-order thermal stratification model can well simulate the temperature distribution in the upper plenum and rapidly elucidate the thermal stratification interface characteristics during the lead-bismuth fast reactor accident. Overall, this study provides an analytical tool for determining the thermal stratification mechanism and reducing thermal stratification.

Density functional calculation of adsorption and dissociation of PbPo molecule on Pd(100), Pd(110) and Pd(111) surfaces

Radiotoxic Po are produced and mainly formed as PbPo in lead-bismuth eutectic in lead-bismuth fast reactor and accelerator-drive systems during normal operation. Part of hazardous PbPo will evaporate and accumulate in the cover gas above the lead bismuth eutectic coolant. To ensure safe handling of the reactor, suitable cover gas filter systems should be fixed but few fundamental data on the interaction between PbPo and filter materials are reported. In this work, adsorption and dissociation of PbPo molecule on candidate filter material Pd surfaces were studied with the density functional theory. It was found PbPo is strongly chemisorbed on Pd(100), Pd(110) and Pd(111) surfaces with adsorption energies in the range of -1.14 eV to -5.36 eV. The activation energies for PbPo dissociation on these tree surfaces are all very small, which are one to two order of magnitude smaller than the recombination activation energies of the reversed paths. The calculated results indicate that strong chemisorption was found between PbPo and perfect low-index Pd surfaces and PbPo molecule is highly inclined to be as Pb and Po atoms on these surfaces.

Conceptual design of oil field energy supply system based on natural circulation lead–bismuth fast reactor

To meet the specific requirements of heavy oil thermal recovery and realize clean and efficient recovery, University of Science and Technology of China (USTC) and China Nuclear Power Technology Research Institute Co., Ltd. jointly designed NCLFR-Oil, an oil field energy supply system based on natural circulation lead–bismuth fast reactor. The system uses a small modular reactor with a thermal power of 40 MWth, which is compact and has a lifetime of more than 15 years. In this paper, the overall design, core neutron characteristics and steady-state thermal–hydraulic design of the reactor are described and evaluated. And the steady-state thermal–hydraulic characteristics of the reactor core as well as the whole three loops are analyzed and discussed. The results show that the reactor meets the safety standards, which means that NCLFR-Oil's conceptual design is acceptable and the reactor has excellent safety characteristics.

Analysis of flow-induced vibration of wire-wrapped fuel assemblies under the liquid metal axial flow in the Gen-IV nuclear reactor

Fluid excitation force is the root cause of flow-induced vibration in the fast reactor core, which is an important cause of nuclear fuel failure and breakage phenomena. In this study, the fluid–structure interaction (FSI) of a filament-wound rod bundles consisting of seven rods in a lead–bismuth fast reactor was simulated. And the fluctuation of fluid excitation force, as well as the vibration displacement and vibration frequency of fuel rods were calculated. It was found that when the fluid flow through the winding wire, the turbulence intensity increases significantly and pressure difference appear on the winding wire windward and leeward surface, which leads to fluctuation in the fluid force, which is the root cause of the emergence of flow-induced vibration. In this structural model, the vibration eventually reach a stable amplitude as well as position, and no pin-to-pin contact occurs.

Thermal design of printed circuit heat exchanger used for lead-bismuth fast reactor

Printed circuit heat exchanger (PCHE) has the potential to replace the traditional heat exchangers in lead–bismuth fast reactors due to the high compactness. In this work, a segmented Log-Mean Temperature Difference (LMTD) method is applied to design a PCHE used for a 280 MWth lead–bismuth fast reactor. The structural strength is checked based on the ASME Code Case. The optimal design of heat exchanger is conducted at different inlet velocities of lead–bismuth eutectic (LBE) and channel diameters. Meanwhile, the variations of LBE-side and water-side Reynolds numbers, convective heat transfer coefficients, mass and volume of the PCHE are analyzed, respectively. The results show that the heat transfer rate per unit volume could reach 181 MW/m3, nearly eight times of shell-and-tube heat exchanger applied to the same lead–bismuth fast reactor, and two times of compact plate heat exchanger with the same working fluids. When the heat transfer rate and the pressure loss are satisfied, the larger the inlet velocity of LBE or the smaller the channel diameter is, the smaller the total volume and mass of the PCHE are.

Effect of surface shot peening on the microstructure and mechanical properties of fuel cladding 1515-Ti austenitic steel

Lead-bismuth fast reactor is considered to be the most promising solution to meet the enormous energy demand in the future. The 15-15Ti austenitic stainless steel with outstanding corrosion and irradiation resistance has been proposed as main cladding candidate material for lead-bismuth fast reactor. However, the mechanical properties of 15-15Ti are weakly influenced by LBE. Shot peening as one of surface strengthening technique can efficiently improve metal mechanical properties. In this paper, shot peening (SP) treatment was applied on 15-15Ti. The microstructures and compositions were investigated using Optical Microscope, Scanning Electron Microscope and X-Ray Diffraction. It concludes that the material hardness is improved, and the residual stress induced by SP did not cause martensitic phase transformation on the sample surface. The presence of slip bands and other obstacles in the reinforcement layer improved the yielding and tensibility of the material.

Study on IMC-PID Control of Once-Through Steam Generator for Small Fast Reactor

The simplification of simulation inevitably leads to model mismatch. In this paper, a once-through steam generator (OTSG) for a small lead bismuth fast reactor (SLBFR) is established and verified, and the OTSG model is simplified by three different methods. Based on the simplified OTSG model, IMC and IMC-PID controllers are designed to verify the sensitivity of the controller to model mismatch. The results show that the sensitivity of the controller to model mismatch is related to the filter parameters. With the increase in λ, the IMC-PID controller becomes insensitive to model mismatch caused by model linearization, non-minimum phase characteristics, noise and time delay. However, the adaptability to model mismatch sacrifices the sensitivity of the system. When λ is too large, the inertia of the controller is too large, resulting in the deterioration of the fast power regulation. Through the research of this paper, the time domain response approximation method is recommended for OTSG model simplification, and λ is recommended to be between 5 and 10 for feedwater IMC-PID controller.

Research on multi-objective optimization method for flow distribution of natural circulation reactor during its life-cycle

A reasonable flow distribution can flatten the temperature distribution at the core outlet and reduce the temperature fluctuation, improving the economy and safety of the reactor. Different from a forced circulation reactor, the flow distribution of a natural circulation reactor is automatically adjusted according to the relationship between the fuel assembly power and inlet resistance. In this work, first, a local optimal flow distribution calculation model is constructed based on the closed parallel multi-channel model. Second, a Multi-objective comprehensive evaluation based on the optimal time zone is proposed to find an optimal flow distribution scheme that satisfies multiple requirements during the entire life-cycle. Third, the flow distribution scheme of a small lead-bismuth fast reactor with long life and natural circulation (SPALLER-100) is studied. The Technique for Order Preference by Similarity to Ideal Solution (TOPSIS) algorithm is chosen as the evaluation method with three evaluating indicators for calculation. Ultimately, the subchannel code is selected to verify the flow distribution scheme. The analysis results show that the scheme based on the power distribution of 3182 days performs the best. Compared with the scheme designed by the beginning of life (BOL), the maximum outlet temperature difference reduced by 30 K, the standard deviation of the maximum outlet temperature difference reduced by 41%, and maximum output power increased by 2.35%. The results are in good agreement with the SUBCHANFLOW code, and the maximum relative error is less than 4.5%.

Intelligent Optimization Method for Core Flow Zoning of Long-Cycle Lead-Bismuth-Cooled Reactor

Flow zoning is an important way to achieve core outlet temperature flattening. Appropriate zoning can improve safety and economy. This study combines an artificial intelligence optimization algorithm with a parallel multi-channel model to develop a model for calculating reactor core flow zoning based on the modern optimization theory, convergence analysis of a genetic algorithm, differential evolution algorithm, and quantum genetic algorithm is carried out for long-life reactor flow partitioning. Using the optimized algorithm, two flow rates are determined using power distribution at the beginning of the core life as the sample data and the maximum power of each fuel assembly during the entire life as the sample data. Comparative analysis of two different flow zoning schemes is implemented on a small long-life natural circulation lead-bismuth fast reactor, SPALLER-100. The findings of this study show that the quantum genetic algorithm has the best convergence for the long-life reactor among the three intelligent optimization algorithms, and it can quickly provide optimal results. In flow zoning scheme calculations based on the core power distribution at the beginning of reactor life, the maximum outlet temperature of the fuel assembly exceeds the thermal safety limit of the reactor, and in the flow zoning scheme calculations based on the average core power distribution during the whole reactor life, the maximum outlet temperature of the fuel assembly is 140 K lower than the maximum outlet temperature obtained in the previous scheme, remaining below the thermal safety limit. The optimal number of partitions for the SPALLER-100 reactor is determined to be 5, and increasing the number of zones only slightly improved the thermal safety performance of the reactor.

Research on Thermal-Hydraulic Parameter Prediction Method of the Small Lead–Bismuth Fast Reactor Core Based on Adaptive RBF Neural Network

In this study, a cladding surface temperature prediction method based on an adaptive RBF neural network was proposed. This method can significantly improve the accuracy and efficiency of the thermal safety evaluation of the lead–bismuth fast reactor. First, based on the sub-channel analysis program SUBCHANFLOW, the core sub-channel model of the small lead–bismuth fast reactor SPALLER-100 was established. Second, the calculated 2000 groups of core power distribution and coolant flow distribution data were used as training samples. The adaptive RBF neural network model was trained to predict the surface temperature of fuel elements in the lead–bismuth fast reactor. Finally, by comparison, the effectiveness and superiority of the adaptive RBF neural network method were proved. The results indicate that the relative error of the maximum temperature of the fuel cladding predicted using the adaptive RBF neural network method was less than 0.5%, which can be used for the rapid prediction of the thermal and hydraulic parameters of the lead–bismuth fast reactor.

Sub-channel analysis of lead–bismuth fast reactor fuel assembly with wire spacers for China initiative accelerator driven system

The lead bismuth eutectic (LBE) cooled fast reactor fuel assemblies in China Initiative Accelerator Driven subcritical System (CiADS) are supported by wire spacers, and the sub-channel code is a significant analysis method to evaluate the thermal–hydraulic behavior of fuel assemblies. In this study, a sub-channel analysis code for LBE coolant is developed, in which the pressure drop model, turbulent mixing model, and heat transfer model based on LBE are added and applied. Besides, the heat transfer experience of the 61-pin fuel assembly and other sub-channel codes are chosen to verify. Finally, the full-scale simulation of CiADS fuel assembly is calculated. The results show that the thermal–hydraulic parameters of CiADS meet the design criterion. The optimized range of pitch to diameter ratio is 1.133–1.145 according to the thermal–hydraulic analysis. It is recommended to use 37-pin bundle for the numerical simulation simplification of full-scale rod bundles.