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In recent times, manufacturing sectors are endeavouring to achieve optimal product quality while minimizing costs. For the accomplishment of this objective, they are focusing on the elimination of the surface finishing and heat treatment processes up to an extent and, opting for concurrent optimization of multiple output responses related to product qualities and machining cost. In machining output optimizations, especially in concurrent optimizations, assigning an equal weightage to all responses leads to substandard results. Keeping in mind, this research work is carried out to eliminate the inadequacy of such optimizations by coupling the multi-response optimization technique Grey Relational Analysis (GRA) with the Analytical Hierarchy Process (AHP) for assigning the different weight fractions to the selected output responses which are tool flank wear, surface roughness and surface microhardness. Face milling experiments on EN-31 steel are conducted engaging hBN nanoparticles-based minimum quantity lubrication system. Employing subjective method-based multi-response optimization, the selected output responses are successfully optimized as per their relative priority and the best parametric conditions are obtained as LFR = 150 ml/h, SD = 40 mm, and 0.5 wt% of hBN nanoparticles in lubricant. Also, it has been noted that the presence of hBN nanoparticles provides effective cooling and lubrication at the cutting zone. [ABSTRACT FROM AUTHOR]

AbstractThe Westinghouse lead-cooled fast reactor (LFR) is a next-generation nuclear power plant whose primary mission is to reduce front-end capital costs and generate flexible and cost-competitive electricity for global markets, while offering mission versatility and satisfying the highest safety and sustainability standards. The LFR is an economical choice for carbon-free power generation to meet the ambitious goals of reducing carbon emissions. The plant utilizes passive safety systems for high reliability and public safety.The LFR testing program aims to fill the technology gaps in the key materials, components, and systems of the LFR plant. It provides demonstrations of LFR engineering that reduce the uncertainty of the LFR design and experimental data for the verification and validation of the modeling and simulation computer codes. With the support of the LFR phenomena identification and ranking table, the phenomena important to plant safety with insufficient states of knowledge have been identified and a Westinghouse testing plan developed to address them.The United Kingdom Advanced Modular Reactor Program Phase 2 was purposed to advance the level of design maturity toward eventual regulatory approval and deployment. Westinghouse and its partners have developed eight state-of-the-art test facilities within the program. Among them, the Passive Heat Removal Facility, Versatile Lead Facility, Lead Water Interaction Facility, and the Lead Freezing Facility are thermal-hydraulic testing facilities, with the other four facilities dedicated to testing lead corrosion and material mechanical behavior in lead.The design, development, and testing of the four thermal-hydraulic facilities are presented in this paper. A variety of computer codes ranging from system-level analysis, component-level analysis, and high-resolution computational fluid dynamics analysis, are utilized to support the development of the facilities. The analyses provide operational and accidental conditions in the LFR to identify major testing conditions and inform the facility design and testing matrix. With a significant amount of experimental data being generated with these highly instrumented test facilities, the computer codes and models will be benchmarked against the experimental results to improve their validation and verification status. [ABSTRACT FROM AUTHOR]

The feasibility of using the Monte Carlo code MCS to generate multigroup cross sections for nodal diffusion simulations RAST-F of liquid metal fast reactors is investigated in this paper. The performance of the MCS/RAST-F code system is assessed using steady-state simulations of the ANTS-100e core. The results show good agreement between MCS/RAST-F and MCS reference solutions, with a keff difference of less than 77 pcm and root-mean-square differences in radial and axial power of less than 0.5% and 0.25%, respectively. Furthermore, the MCS/RAST-F reactivity feedback coefficients are within three standard deviations of the MCS coefficients. To validate the internal thermal-hydraulic (TH) feedback capability in RAST-F code, the coupled neutronic/TH1D simulation of ANTS-100e is performed using the case matrix obtained from MCS branch calculations. The results are compared to those obtained using the MARS-LBE system code and show good agreement with relative temperature differences in fuel and coolant of less than 0.8%. This study demonstrates that the MCS/RAST-F code system can produce accurate results for core steady-state neutronic calculations and for coupled neutronic/TH simulations.

SUMMARY: Switch defective/sucrose non‐fermentable (SWI/SNF) chromatin remodeling complexes are evolutionarily conserved, multi‐subunit machinery that play vital roles in the regulation of gene expression by controlling nucleosome positioning and occupancy. However, little is known about the subunit composition of SPLAYED (SYD)‐containing SWI/SNF complexes in plants. Here, we show that the Arabidopsis thaliana Leaf and Flower Related (LFR) is a subunit of SYD‐containing SWI/SNF complexes. LFR interacts directly with multiple SWI/SNF subunits, including the catalytic ATPase subunit SYD, in vitro and in vivo. Phenotypic analyses of lfr‐2 mutant flowers revealed that LFR is important for proper filament and pistil development, resembling the function of SYD. Transcriptome profiling revealed that LFR and SYD shared a subset of co‐regulated genes. We further demonstrate that the LFR and SYD interdependently activate the transcription of AGAMOUS (AG), a C‐class floral organ identity gene, by regulating the occupation of nucleosome, chromatin loop, histone modification, and Pol II enrichment on the AG locus. Furthermore, the chromosome conformation capture (3C) assay revealed that the gene loop at AG locus is negatively correlated with the AG expression level, and LFR‐SYD was functional to demolish the AG chromatin loop to promote its transcription. Collectively, these results provide insight into the molecular mechanism of the Arabidopsis SYD‐SWI/SNF complex in the control of higher chromatin conformation of the floral identity gene essential to plant reproductive organ development. Significance Statement: This study provides insight into the molecular mechanism of the Arabidopsis SYD‐SWI/SNF complex in the control of higher chromatin conformation of the floral identity gene essential to plant reproductive organ development. [ABSTRACT FROM AUTHOR]

With the rise of economic and safety standards for nuclear reactors, new concepts of Gen-IV reactors and modular reactors showed more complex designs that challenge current tools for reactor physics analysis. A Monte Carlo (MC) two-step method was proposed in this work. This calculation scheme uses the continuous-energy MC method to generate multi-group cross-sections from heterogeneous models. The multi-group MC method, which can adapt locally-heterogeneous models, is used in the core calculation step. This calculation scheme is verified using a Gen-IV modular lead-based fast reactor (LFR) benchmark case. The influence of homogenized patterns, scatter approximations, flux separable approximation, and local heterogeneity in core calculation on simulation results are investigated. Results showed that the cross-sections generated using the 3D assembly model with a locally heterogeneous representation of control rods lead to an accurate estimation with less than 270 pcm bias in core reactivity, 0.5% bias in control rod worth, and 1.5% bias on power distribution. The study verified the applicability of multi-group cross-sections generated with the MC method for LFR analysis. The study also proved the feasibility of multi-group MC in core calculation with local heterogeneity, which saves 85% time compared to the continuous-energy MC.

A Lead-Cooled Fast Reactor (LFR) is one of the six advanced fourth-generation reactors that many countries have been interested in designing, constructing, or develop them in the world. ALFRED reactor design investigated in this paper has 125 (MW) electric power and is being cooled by lead. The main purpose of the ALFRED project is to design and construct a reactor with high safety performance and reliability in all operation modes. In the present study, Monte Carlo codes MCNPX and SuperMC are used to calculate the neutronic parameters of the ALFRED reactor core including effective multiplication factor, kinetic parameters, neutron flux, and power distribution based on the latest design data. According to the results, the insertion of the control rods into the core can move the maximum flux height of about 40% to 50% to the end. Also, comparing codes' results show an appropriate consistency. [ABSTRACT FROM AUTHOR]

Single-point incremental forming process is a cost-effective die-less process to form sheet metal parts. To implement the forming method in industrial applications, it is required to study forming of various shapes. The available literature on hole flanging by single-point incremental forming is mostly focused on cylindrical flanges and very few studies on the square flange are available. The available study on the square flange is based on multistage experiments. The single-stage forming can reduce the forming time significantly. Therefore, in this work, the single-stage single-point square hole flanging of AA5052-H32 sheet material has been studied. A process-specific fracture forming limit line is obtained by single point incremental hole flanging experiments to analyze the strain behavior during the forming of flanges. A definition of Limit Forming Ratio for square flanging has been proposed and its value for the flange forming has been obtained experimentally. The effect of precut square hole length, pre-cut square hole corner radius, and tool rotation on flange formability has been studied and presented. Moreover, forming forces evolving during the process, the microstructure of the material, and the surface roughness of the flange are analyzed and presented. The results show that the flange formability increases with increase in the pre-cut square hole corner radius; the tool rotation reduces the flange formability and increases the surface roughness. [ABSTRACT FROM AUTHOR]

Low temperatures restrict the growth and geographic distribution of plants, as well as crop yields. Appropriate transcriptional regulation is critical for cold acclimation in plants. In this study, we found that the mutation of Leaf and flower related (LFR), a component of SWI/SNF chromatin remodeling complex (CRC) important for transcriptional regulation in Arabidopsis (Arabidopsis thaliana), resulted in hypersensitivity to freezing stress in plants with or without cold acclimation, and this defect was successfully complemented by LFR. The expression levels of CBFs and COR genes in cold-treated lfr-1 mutant plants were lower than those in wildtype plants. Furthermore, LFR was found to interact directly with ICE1 in yeast and plants. Consistent with this, LFR was able to directly bind to the promoter region of CBF3, a direct target of ICE1. LFR was also able to bind to ICE1 chromatin and was required for ICE1 transcription. Together, these results demonstrate that LFR interacts directly with ICE1 and activates ICE1 and CBF3 gene expression in response to cold stress. Our work enhances our understanding of the epigenetic regulation of cold responses in plants. [ABSTRACT FROM AUTHOR]

Given the ever-increasing global demand for energy and the need to reduce greenhouse gas emissions, small modular reactors (SMRs), have emerged as potential options for increasing the contribution of nuclear energy, offering lower costs and faster deployment compared to traditional nuclear projects. In the context of this technological development, safety studies have become a priority, particularly for licensing new-generation systems such as metal-cooled fast reactors. This work models the steady-state operation of the lead-cooled SMR SEALER Arctic using Computational Fluid Dynamics. The entire primary circuit of the SEALER is modeled; the core is represented as a combination of porous media and heat sources, the pumps are represented as recirculating boundary conditions to account for momentum sources, and the steam generators are represented as porous media coupled with a temperature-dependent heat sink function. The main objective of this study is to simulate the SEALER under steady-state condition, while also accounting for the effects of heat conduction through its solid regions, and heat losses on the reactor vessel wall to the environment. For the former, the reactor is modeled with and without conductive solids and surfaces, using a conjugate heat transfer model. For the latter, natural convection and radiation heat transfer considerations are included as boundary conditions, and a parametric study is carried out with a range of external temperatures, and their effects on fuel and coolant temperatures are also discussed. Despite significant differences in local temperatures near the vessel walls, the impact on the peak fuel temperature and the average coolant temperature was less noticeable. Ultimately, the general operating parameters of the steady-state reactor design were verified, which is the first step before using the current model to evaluate fast transients and postulated events, where the thermal inertia of the solids and additional heat losses could play a crucial role on determining the system's response to rapid temperature changes. • CHT modeling enhances visualization of the temperature fields inside lead-cooled reactors. • CHT has little impact on the reactor's volume-average temperature in steady-state. • Heat losses through the outer wall leads to relevant local temperature drops. • Fuel peak temperature is weakly influenced by CHT and heat losses to the domain's outside. [ABSTRACT FROM AUTHOR]

• Refined subchannel analysis code is developed for lead-based wire-wrapped bundles. • A refined flow mixing module is applied to improve the prediction accuracy of crossflow distribution. • A refined fuel rod model is applied for circumferential non-uniform heat transfer. • The code is applied in the thermal–hydraulic analysis of a non-uniformly heated fuel assembly. Lead-based fast reactor (LFR) has been considered as one of the most promising concepts for Generation IV nuclear reactors. Subchannel analysis is frequently used to obtain the flow field and temperature distribution. Its applicability in lead-based wire-wrapped bundles has to be improved due to complex thermal-hydraulic characteristics. In this study, a refined subchannel analysis code, STAR-LFR, is developed. A refined fuel rod model is applied for circumferential non-uniform heat transfer, while a refined flow mixing module is applied for flow sweeping caused by wire spacers. Code validation with respect to crossflow distribution and heat transfer are conducted using high-fidelity large eddy simulation and experimental data, respectively. Besides, the code is applied in the thermal–hydraulic analysis of a non-uniformly heated fuel assembly. Overall, great ability on flow and heat transfer prediction are proved. The crossflow rate and the overheat caused by wire can be well estimated. This work provides a reliable tool that could support the thermal–hydraulic analysis as well as the design of LFR. [ABSTRACT FROM AUTHOR]

[Display omitted] • Various crack morphologies are observed on T91 steels after exposure to stagnant oxygen-saturated LBE at 500 °C. • A thermo-mechanical peridynamic model is introduced for cooling-induced fracture in the oxide scale. • Temperature-dependence of thermo-mechanical properties in the duplex oxide scale and porosity in the outer layer are considered. • Model reproduces the cracks' morphologies and spacing. Ensuring the continuity and stability of the oxide scale to separate T91 steel from lead-bismuth eutectic (LBE) is an effective method for limiting corrosion in the lead-based fast reactor system. In this study, we specifically investigated the impact of cooling thermal stress on the stability and damage of the oxide scales of T91 steel when exposed to oxygen-saturated liquid lead–bismuth eutectic at high temperatures. A corrosion test was conducted on T91 steels exposed to stagnant oxygen-saturated LBE at 500 °C. Experimental results indicated that oxide damage exists without any external force, including interface cracks and through-scale cracks perpendicular to the interface. We developed a three-layer peridynamic model of T91-(Fe,Cr) 3 O 4 -Fe 3 O 4 to simulate the deformation and failure of oxides under a cooling process from high temperature to room temperature. The simulation results show that a temperature drop of about 200 °C from 500 °C can cause through-oxide cracks in the oxide scale, and subsequent cooling can lead to the propagation of interfacial cracks. Additional modifications were introduced in the model to account for the porosity of the oxide scale. Parametric investigations illustrate the effects of oxide scale thickness and porosity on the resulting crack patterns. [ABSTRACT FROM AUTHOR]

This paper reviews the research progress of LBE corrosion-resistant coatings based on the conducted experiments in the published literature. Metallic coatings have been widely studied due to their outstanding bonding properties with steel substrates. Different metallic coatings exhibit different corrosion resistance, which is attributed to the formation of a dense and continuous protective oxide scale during corrosion. The ceramic coatings prepared by different methods show different microstructure and corrosion resistance. The attempt at composite/multilayer coating strategy provides a new idea for obtaining coatings with specific properties. Different coatings and coatings prepared by different methods have different properties after irradiation. The development of simultaneous proton irradiation and LBE corrosion coupling experiments provides a new way for the screening of coatings with better comprehensive performance under near LFR service conditions. The summary of the existing LBE corrosion-resistant coatings in this review will help the authors to prepare and screen coatings with excellent performance. • The research progress of the LBE corrosion-resistant coatings was reviewed. • Different coatings and different preparation processes show different corrosion resistance. • The defects introduced by irradiation deteriorate the corrosion resistance of the coatings. • The simultaneous irradiation-corrosion device provides a new choice to evaluate the performance of the coatings. [ABSTRACT FROM AUTHOR]

• An inhouse STH code named SACLER is modified developed with LBE and pure lead thermal properties and heat transfer models. • The liquid metal Helical Coiled Once Through Steam Generator model is applied in SACLER code. • Model separation validation and system integrity validation are carried out for the code. LFRs are one of six advanced Generation IV reactor concepts using lead-based coolant with high thermal conductivity, and HCOTSGs are generally adopted to highlight LFR miniaturization characteristics. An inhouse STH code SACLER with the HCOTSG model and the cover gas model was developed specifically for LFRs. NACIE facility was used to demonstrate SACLER accuracy. The HCOTSG model was verified using the RELAP code, experimental data, and ELSY design parameters. The validation results are consistent with the experimental data of NACIE. The HCOTSG model calculation results are in good agreement with the experimental data and the RELAP code. The error between the ELSY simulation result and the design value is within 1.0%. Liquid level simulation of ELSY shows that liquid levels are mainly determined by the pump head and flow resistance. SACLER can accurately simulate the LFR system with HCOTSG, including transient and steady-state conditions. [ABSTRACT FROM AUTHOR]

The Lead-bismuth cooled Fast Reactors (LFR) utilizes wrapped hexagonal fuel assemblies arranged in a honeycomb pattern. The inter-wrapper flow (IWF) between adjacent assemblies has a significant influence on the core thermal-hydraulics. In this paper, the LFR core assembly analysis model, inter-wrapper flow and heat transfer model, and multi-scale coupling analysis model were developed and were implemented into the previously self-developed subchannel-level three-dimensional thermal hydraulic analysis code CorTAF-LBE for whole core of LFR, which is based on the OpenFOAM computational fluid dynamics simulation platform. The inter-wrapper flow and heat transfer model was validated against the KALLA-IWF experiment. Whole core steady state operating condition and low flow rate condition simulations were performed with reference to the 19-rod bundle geometry from KALLA laboratory and typical LFR core assembly arrangement. The whole core steady state distribution laws of key thermal-hydraulic parameters were obtained, and the influence of low flow rate operation on safety parameters such as cladding and pellet temperatures was analyzed. Under low flow rate operating conditions, the IWF heat transfer was accounted for 4.87% of the core thermal power, which was 315.6 kW higher than that under normal operating conditions, playing an important role in core heat removal and safe operation. This work provides important insights into LFR core design and the assessment of the role of IWF under low flow rate conditions. • The Multi-scale coupling analysis model is established and implanted into the CorTAF-LBE code. • The steady-state thermal-hydraulic parameter distribution at whole core scales is analyzed. • The effect of inter-wrapper flow on core heat transfer at low flow rate condition is evaluated. [ABSTRACT FROM AUTHOR]

The thermal inertia results in net heat absorption in natural circulation lead-cooled reactors during power variation processes, increasing the risk of core meltdown. In this study, a dynamic simulation model of a natural circulation lead-cooled fast reactor coupled with a supercritical carbon dioxide Brayton Cycle power generation system has been developed based on Vpower. The influence of thermal inertia on system operating temperature under different power variation rates was discussed. The results indicate that during power reduction, different decline rates only affect the temperature variation range of the system without impacting the final stable state. However, during power increase, different ramp rates not only affect the temperature variation range but also lead to varying degrees of system temperature rise. To address this issue, an optimized turbine throttling scheme is proposed. Compared to the original plan, the new plan successfully stabilizes the temperature at the rated value after increasing the system power. This study provides insights for the operation control strategy of similar reactor types. [ABSTRACT FROM AUTHOR]

The current study is mainly focused on the monthly variation in the lightning flash rate (LFR) and related thermodynamic parameters using the data for the years 2000–2013, and the trend of lightning variation is explored. Lightning data are used from a lightning imaging sensor (LIS) and an optical transient detector (OTP) boarded on the tropical rainfall measuring mission (TRMM). Additionally, aerosol optical depth (AOD) data at 550 nm for the same period were considered from a Moderate Resolution Imaging Spectroradiometer (MODIS). The assessment of lightning and AOD using monthly data makes it difficult to study seasonal contributions, and higher-resolution (hourly) data may be more appropriate, but unfortunately, no data were available with a higher resolution than monthly. The dependency of LFR is also investigated using thermodynamic/dynamic parameters. The LFR shows a moderate correlation with a correlation coefficient of 0.56, 0.62, and 0.63 for AOD, CAPE, and vertical velocity, respectively. The increasing AOD in the pre-monsoon season is associated with higher lightning flash rates over this region. The possible sources of aerosols that cause an increase in lightning activities are identified from the classification of aerosols based on the characteristic values of the AOD and the Ångström exponent. The thermodynamic relation of the Product of Bowen ratio with the sum of the precipitation rate and evaporation rate has been used as a proxy to evaluate the lightning flash rate density over Srinagar, Uttarakhand region (78.55° E–79.05° E, 29.97° N–30.47° N), with nine models from the Coupled Model Inter-comparison Project-Phase 5 (CMIP5). The model-simulated LFR has also been used for the projection of lightning in the late 21st century, and the projected LFR over the study area shows a 7.41% increase during the (2079–2088) period as compared to the historic period (1996–2005). The results of the study region indicate caution in using any single climate variable as a proxy for projecting a change in the lightning–climate relationships in the scenario of global warming. [ABSTRACT FROM AUTHOR]

BackgroundThe key factors affecting the thermal performance of the lead-cooled fast reactor (LFR) annular fuel are very important to improve the core power density of LFR.PurposeThis study aims to analyze those key factors and provide guidance for improving reactor core power density.MethodsFirstly, one quarter thermal analysis model of typical annular fuel element of lead cooled fast reactor was established, followed by the thermal resistance analysis of typical LFR annular fuel element to evaluate the key factors affecting its thermal performance. Secondly, four indicators, i.e., the temperature difference between the coolant outlet inside and outside the channel, the maximum temperature of annular fuel pellet, the position of adiabatic surface and the temperature difference between the inner and outer cladding, were used as evaluation criteria to analyze the influence of various factors on the thermal performance of the annular fuel element.ResultsThe thermal resistance analysis results show that the key factors affecting the thermal performance of LFR annular fuel element are gas gap width, cladding thickness and flow distribution ratio. Evaluation and analysis results indicate that temperature difference of coolant and the maximum temperature of annular fuel pellet can be significantly reduced by increasing the flow distribution ratio whilst slightly increasing the width of the outer gas gap and reducing the width of the inner gas gap can reduce the coolant outlet temperature difference and the average temperature difference between the inner and outer cladding. Changing the cladding thickness has no significant effect on the thermal performance of annular fuel element.ConclusionsStudying the key factors that affect the thermal performance of the LFR annular fuel has importance value for optimal design of the LFR core.

Accurate measurements of the oxygen concentration in liquid lead-bismuth eutectic (LBE) alloys are essential for ensuring optimum cooling conditions in heavy liquid metal coolant fast reactors. However, while unavoidable variations in the Bi contents of these alloys can affect the accuracy of these measurements, this issue has not yet been rigorously evaluated. The present work addresses this issue by evaluating variations in oxygen concentration measurements obtained experimentally for oxygen-saturated liquid LBE alloys with different Bi contents at different temperatures using a solid electrolyte oxygen sensor. Theoretical relationships are established between sensor voltage, oxygen concentration, temperature, and Bi content under the applied experimental conditions, and excellent agreement is obtained between the experimental and theoretical results. The theoretical relationships establish that the voltage decreases by about 1 mV for every 1 wt% increase in Bi content at a temperature of 400 °C. At a fixed voltage, with the increase of T, the influence of the Bi content on the CO calculation results gradually increased. Finally, the impact of varying Bi contents on the accuracy of oxygen concentration measurements is evaluated under practical reactor operation conditions, and higher requirements for the control of dissolved oxygen in LBE-cooled fast reactors are proposed.

Lead alloy is used as coolant in Lead-based cooled Fast Reactor (LFR). The natural characteristics of lead alloy are combined with the simple structural design of LFR. This constitutes the inherent safety characteristics of LFR. The main work of this paper is to take the main coolant pump (MCP) in the lead-cooled fast reactor (LFR) as the research object, and to study the flow pattern distribution of the internal flow field under the reverse rotation pump condition, the reverse rotation positive-flow braking condition and the reverse rotation negative-flow braking condition. In this paper, the double-outlet volute type and the space guide vane are selected as the potential designs of the CLEAR-I MCP. In this paper, the CFD method is used to study the potential reverse accident of the MCP. It is found that the highest flow velocity in the impeller appears at the impeller outlet, and the Q-H curves of the two design programs basically coincide. The space guide vane type MCP has better hydraulic performance under the reverse rotation positive-flow condition, the Q-H curves of the two designs gradually separate with increasing flow rate, and the maximum flow velocity inside the space guide vane type MCP is obviously lower than that of the double-outlet volute type. For the reverse rotation test of MCP, only the condition of the forward rotating pump of the main coolant pump is tested and verified. For the simulation of the MCP in LBE medium, it proved that the turbulence model and basic settings selected in the simulation are reliable.

• SPARK-NC is a lead-bismuth cooled fast reactor with promising safety features. • Investigation on radionuclide inventory evolution performed using ORIGEN2.2 and NECP-SARAX. • Reactor specific one group neutron cross-sections library generated for ORIGEN2.2. • Radionuclides inventory reported in this study by the two code systems agree within 10%. • Central core assemblies have ∼25% higher inventory than average value. SPARK-NC, a 10 MW(e) lead–bismuth eutectic (LBE) cooled fast reactor design, exhibits promising characteristics like inherent gamma shielding, natural circulation, and a high boiling point. Following detailed neutronic studies, a thorough investigation of nuclear safety necessitates a detailed analysis of the core radionuclide inventory. This information is particularly crucial for source term calculations, which play a vital role in assessing the potential radiological consequences. This study establishes the life-cycle inventory of SPARK-NC using two independent computational systems: ORIGEN2.2 and NECP-SARAX. ORIGEN2.2, equipped with a reactor-specific library generated by NECP-MCX, is used for average whole-core inventory analysis. NECP-SARAX, on the other hand, explicitly considers core heterogeneity in terms of enrichment, specific power, and burn-up. This work presents the radionuclide inventories and the relative calculation differences observed between the codes. Actinides like uranium and curium display minimal code dependence, while plutonium isotopes exhibit a maximum relative difference of 8 %. Fission products generally agree within 5 %, except for I-131, which shows a discrepancy of around 10 %. The activity of I-131 and Cs-137 are estimated to be approximately 1 × 1016 Bq and 3 × 1015 Bq, respectively. Additionally, the photon source strength is 1017/s at 1 MeV, dropping to 1016/s below 6 MeV. Fission products and actinides contribute a decay heat of 0.65 MW. Assembly-wise analysis reveals a direct proportionality between radionuclide inventory and peaking factor, with the average assembly inventory being roughly 25 % lower than the peak assembly inventory. Rare earth elements (Ce, Sm, Pm, Pr, Nd, La, Y) exhibit a maximum mass of approximately 8.5 kg with a 3 % relative difference between the codes. Conversely, halogens (I, Br) have a minimum mass of around 0.2 kg with a 13 % relative difference. These findings, alongside the quantification of radionuclides, provide valuable insights into the code selection for future analyses of SPARK-NC and similar reactor systems. [ABSTRACT FROM AUTHOR]

• An overview of the ATHENA experimental facility is provided. • A numerical model of ATHENA has been realized with RELAP5/Mod3.3, properly modified to include heavy liquid metals. • A Loss-Of-Heat-Sink transient has been simulated with RELAP5/Mod3.3. • A sensitivity analysis on the Main Heat Exchanger outlet temperature set point for the shutdown has been performed. ATHENA (Advanced Thermal-Hydraulic Experiment for Nuclear Applications) is a large multipurpose pool-type lead-cooled facility under construction at the Mioveni site in Romania. It has been identified by the FALCON (Fostering ALfred CONstruction) Consortium to characterize large to full-scale ALFRED components, to conduct integral tests, and to investigate the main thermal–hydraulic phenomena inherent in pool-type systems. ATHENA is representative of ALFRED in terms of the difference in height of the thermal barycenters of the heat source and heat sink, i.e., 3.3 m, in order to reproduce the buoyancy forces in the system. Similar to ALFRED's design, ATHENA minimizes thermal stratification within the main vessel even under natural circulation conditions, through an internal structure referred to as "barrel". This structure directs the fluid flow towards the main vessel, preventing fluid stagnation near the vessel itself. The paper initially provides a steady-state thermal–hydraulic characterization of the facility, including details of the numerical model developed using the RELAP5/Mod3.3 thermal–hydraulic code. Then, focus is given to the transient analysis considering as a reference scenario a Loss-of-Heat-Sink (LOHS) accidental transient. In this scenario, the Main Circulation Pump (MCP) is assumed to remain operational while the Core Simulator (CS) is deactivated once the lead temperature at the Main Heat Exchanger (MHX) outlet reaches a predefined threshold. A sensitivity analysis is conducted with set points of 430 °C, 450 °C, 470 °C, and 490 °C, assessing the system's response following MHX isolation from the secondary loop. The study evaluates the impact of different CS deactivation set points on reactor SCRAM delay (reducing CS power to a level representative of decay heat) as well as on system maximum and minimum temperatures. [ABSTRACT FROM AUTHOR]

Blockage accidents are critical scenarios in the design and safety analysis of lead-bismuth cooled fast reactor (LFR) core. Traditional analysis of blockage accidents in LFR focuses on localized, fine-scale computational fluid dynamics (CFD) simulations of single or three assemblies, but the analysis of the whole core scale impact caused by blockage accidents is insufficient. Therefore, this paper uses CorTAF-LBE, a three-dimensional thermal-hydraulic analysis code developed by XJTU-NuTHeL, to analyze the impact of blockage accidents on the whole core of the LFR. The reliability of the code in calculating thermal-hydraulic parameters under blockage accidents was validated based on the KALLA-THEADES and KALLA-IWF experiments. Taking the MYRRHA-FASTEF core as the object, simulations and analyses are conducted for various blockage scenarios with different lengths and positions. The results indicate that blockage accidents lead to an enlarged coolant temperature gradient at the core outlet. Lengthening the blockage results in an elevation of the peak temperature in the cladding. Under 2.06 % blockage at the center of the assembly, blockage in the middle of the heating segment poses the greatest threat to cladding integrity, with the maximum temperature reaching 1336.9K, an increase of 635.4K compared to normal operating conditions. Under 4.59 % blockage at the edge of the assembly, the maximum cladding temperature reaches 1381.8K, and the heat transfer rate of the inter-wrapper flow (IWF) adjacent to the blockage area is 24.2 % higher than under normal operation. Additionally, severe degradation in heat transfer downstream was not observed in several simulated blockage scenarios. [ABSTRACT FROM AUTHOR]

• The experimental data of energy release and debris fragmentation characteristics during MFCI process of SS-LBE has been obtained. • The fragmentation mechanism is discussed from hydraulic and thermal aspects to further understand the behavior of molten materials in LBE pool. • The relationship between the cumulative mass percentage of SS fragments and the fragment size approximately satisfies a power function, and the power exponent lies between 0.7 and 1.7. • Simplified models of thermal fragmentation have been established, where the innermost layer of the solidified shell is subjected to the maximum tangential compressive stress, while the outermost layer to the maximum tangential tensile stress. • Comparing the existing models with models proposed in this paper. The LFR is a candidate for the Generation-IV nuclear systems, where the process of MFCI differs from that occurred in LWRs and SFRs. In the absence of experimental study and widely applied theoretical models for LFRs, experiments on the interaction between molten SS and liquid LBE are conducted in this study under eight conditions. The morphological characteristics and fragment sizes of the products are analyzed based on the hydrodynamic effect parameter (We) and thermal effect parameter (T ic). Simplified thermal fragmentation models are established, where the innermost layer of the solidified shell is subjected to the maximum tangential compressive stress, while the outermost layer is subjected to the maximum tangential tensile stress. As the shell thickness increases, the total tangential stress of the outermost layer decreases. The fragmentation will occur when the maximum tangential stress on the shell exceeds the shell tensile strength. [ABSTRACT FROM AUTHOR]

Due to the complex structure of the LFR fuel rod bundle, refined CFD simulation consumes significant computing resources and time, making it difficult to meet the immediate needs of engineering design and research. This paper introduces a method that integrates the sub-channel level thermal-hydraulic analysis code CorTAF-LBE with the BPNN algorithm to analyze the thermal-hydraulic phenomena within the LFR core. The method has been validated by refined simulation results of the THEADES test section, which indicate that the error is within an acceptable range. The feasibility of its application in LFR fuel assemblies was also evaluated. The flow characteristics and temperature distribution in unwrapped LFR rod bundles are investigated. Due to differences in fuel distribution and flow area, the temperature in the central channel is much higher than that in the side and trapezoidal channels. Three accident conditions, including inlet blockage, center blockage, and outlet blockage, were simulated. The corresponding peak temperatures of the fuel cladding were 670 K, 767 K, and 860 K, respectively, representing an increase of approximately 170 K compared to the nominal conditions. The analysis results effectively demonstrate the 3D temperature distribution of the fluid in the rod bundles, which is significant for reactor design and safety analysis of LFRs. [ABSTRACT FROM AUTHOR]

The China initiative Accelerator Driven System (CiADS) is a 10 MW Lead-Bismuth-Eutectic (LBE) cooled sub-critical Fast Reactor (LFR) driven by a superconducting Linac coupling with a LBE spallation target. To operate CiADS from low power to full thermal power in 10 MW, it is necessary to develop a wide-range neutron flux monitor (NFM) for power monitoring. To cover over 7 order neutron flux, a fission chamber (FC) would be the best choice. This work focuses on the in-core neutron monitoring by using U-235 micro fission chamber (MFC) for study on local power variation. In addition, in-core fission rate can be obtained with MFC neutron measurement. CiADS-like core of LFR is built with a measurement channel to place MFC in the response simulations. The core k eff is changed within 100 pcm with a measurement channel moving from center to outward reflector in different assembly circles. Neutron proportions in three energy regions, mean neutron energy, and fission/capture reaction rate and their ratios are almost the flat in the axial distribution of fuel assemblies. There is a peak of mean neutron energy and high energy neutron component around target in the axial distribution of the first LBE assembly which is between target and fuel assembly. Radial distributions of high and intermediate energy neutron proportions are almost flat, but that of low energy neutron proportions is changed over 5 times. Neutron sensitivities of different detectors are calculated by using 3 groups neutron to fit the thermal neutron sensitivity provided by different manufacturers. Radial distributions of neutron responses are almost flat in the axial center positions of each assembly, but changed observably in outward axial positions. According to the simulations of in-core neutron response, it is demonstrated that detectors can work on the Pulse mode and Campbell mode to cover dynamic range from 1 mW to 20 MW in the neutron monitoring of CiADS-like core. [ABSTRACT FROM AUTHOR]

The 9–12% Cr ferritic–martensitic heat-resistant steel is the main candidate structural material for the Lead-cooled Faster Reactor. The lower Gibbs free energy change of Si oxide can promote the formation of a stable oxide layer, which can improve the corrosion resistance of the material. Therefore, it is of great significance to study the effect of silicon (Si) on the corrosion resistance of T91 steel in lead–bismuth eutectic (LBE). The corrosion resistance of T91 steel with Si contents of 0.5 wt.%, 1.3 wt.%, and 2.0 wt.%, both in dynamic and static LBE at 550 °C, was investigated. The microstructure was analyzed by X-ray diffraction (XRD), scanning electron microscope (SEM) and transmission electron microscope (TEM), while the oxide films were characterized by electron probe microanalysis (EPMA). Results show that the addition of Si is conducive to improving the corrosion resistance of T91 steel in LBE. T91 steel with high Si content has a thinner and more stable oxide film. The change of Si content can change the contact angle between the steel and LBE, and the contact angle is the largest when the Si content is 1.3 wt.%. The Si-rich oxide layer is usually located in the inner oxide layer, which promotes the formation of a Cr oxide layer located in the internal oxidation zone (IOZ). Si will not enter the precipitated phase, but only change the ferrite content. The oxidation model of T91 steel containing Si in LBE was also proposed. [ABSTRACT FROM AUTHOR]

Small Modular Reactors (SMRs) are attracting wide attention due to their outstanding performance, extensive studies have been carried out for lead-based fast reactors (LFRs) that cooled with Lead or Lead-bismuth (LBE), and small modular natural circulation LFR is one of the promising candidates for SMRs and LFRs development. One of the challenges for the design small modular natural circulation LFR is to master the natural circulation thermal-hydraulic performance in the reactor primary circuit, while the natural circulation characteristics is a coupled thermal-hydraulic problem of the core thermal power, the primary loop layout and the operating state of secondary cooling system etc. Thus, accurate predicting the natural circulation LFRs thermal-hydraulic features are highly required for conducting reactor operating condition evaluate and Thermal hydraulic design optimization. In this study, a thermal-hydraulic analysis code is developed for small modular natural circulation LFRs, which is based on several mathematical models for natural circulation originally. A small modular natural circulation LBE cooled fast reactor named URANUS developed by Korea is chosen to assess the code's capability. Comparisons are performed to demonstrate the accuracy of the code by the calculation results of MARS, and the key thermal-hydraulic parameters agree fairly well with the MARS ones. As a typical application case, steady-state analyses were conducted to have an assessment of thermal-hydraulic behavior under nominal condition, and several parameters affecting natural circulation were evaluated. What's more, two characteristics parameters that used to analyze natural circulation LFRs natural circulation capacity were established. The analyses show that the core thermal power, thermal center difference and flow resistance is the main factors affecting the reactor natural circulation. Improving the core thermal power, increasing the thermal center difference and decreasing the flow resistance can significantly increase the reactor mass flow rate. Characteristics parameters can be used to quickly evaluate the natural circulation capacity of natural circulation LFR under normal operating conditions.

The lead-cooled fast reactor (LFR) is one of six reactor concepts selected in the Generation IV Technology Roadmap and is perhaps the first to be applied commercially. Because the heavy liquid metal coolant has a severe corrosion effect on the core structure, the compatibility of the heavy liquid metal coolant and structural materials is recognized as a key limitation in the design and application of the LFR. Corrosion by heavy liquid metals such as liquid lead or lead–bismuth eutectic (LBE) is a physical or physical–chemical process involving surface oxidation, dissolution of material constituents, erosion corrosion, and fretting corrosion. Corrosion by heavy liquid metal can change the microstructure, composition, and surface morphology of structural materials, which will affect their mechanical and physical properties and lead to system failure. Currently, LFR research institutes are devoting great effort to the research and development of structural materials with good high-temperature mechanical properties and excellent corrosion and irradiation resistances. In this study, a series of experiments and analyses were performed on self-developed 11Cr−1Si ferritic/martensitic (F/M) steel, including heat treatment tests, mechanical tests, corrosion tests in static lead-bismuth eutectic (LBE), and slow strain-rate tests (SSRT) in LBE. The heat treatment results show that 11Cr−1Si steel obtains a good combination of high strength and toughness after quenching at 950 ℃ and tempering at 750 ℃. 11Cr−1Si steel was found to have good LBE corrosion resistance after exposure in static LBE for 3368 h, with a sufficiently low oxidation rate and a continuous and compact surface oxide layer, which protect the base metal of 11Cr−1Si from LBE penetration. The SSRT results show that the ductility of 11Cr−1Si in contact with LBE is sensitive to temperature, with loss of ductility observed at 350 ℃ and 400 ℃, but not at 450 ℃.

In current context, to achieve lower heat loss in Linear Fresnel reflector (LFR) unit, innovative configuration for cavity has been offered. This cavity consists of one hot circular wall and two inclined adiabatic walls. The below wall is made of glass and connect with outside air flow. The operating fluid is air and its features were temperature dependent. Inclination angle (θ), wind speed and circular wall temperature are variable factors. Both thermal and irreversibility behavior were analyzed in outputs. Finite volume method (FVM) was utilized for such 2D problem involving radiation source term. The incident radiation contours were reported which were obtained by implement of discrete ordinates (DO) model. Verification was done with comparing data with that of previous numerical publication. Stream function augments about 87.5% with augmentation of θ while changing the hot temperature cannot affect the strength of the eddy. Radiation mode is directly relevant to temperature difference, so lower glass temperature provides lower incident radiation. At lowest Th, augment of θ offers reduction in glass temperature and incident radiation about 2.27% and 12.94%. Given θ = 120o, reduce of Tr makes cold temperature to increase about 20.59%. Owing to augmentation of heat loss with rise of Th, θ and wind speed; all factors have direct relation with Nu and Nur. Temperature of circular wall has highest influence on Nu. With assumption of minimum values of other factors, Nu augments 19.63% and 65.07% with rise of hext and θ while it declines about 89.59% with increase of Tr. Irreversibility decreases about 86.38% with rise of Tr when θ = 120o, hext = 20. Given Th = 673 K, hext = 20, total entropy augments about 6.75% with employing greater angle. [ABSTRACT FROM AUTHOR]

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The current study proposes a novel solar PV and thermal (PVT) system based on the combination of linear Fresnel reflector (LFR) concentrator and ITO-EG nano-fluids optical filter. Preparation and relevant measurements of the ITO nano-particles as well as the ITO-EG nano-fluids are conducted. The test results reveal that the optical filter has an average absorptivity of 30.9% in the spectrum range of 250.0–2500.0 nm. The solar concentration behaviour of the PVT system is revealed and the sensitivity analysis results of sun tracking error indicate that when the north-south tracking error is 0.2°, the overall optical efficiency is 90.12%, which means the PVT system has a relatively good adaptive faculty on sun tracking error. Theoretical thermodynamic calculation and CFD simulation methods are utilized to estimate the operation behaviour of the PVT system. The results demonstrate that the PVT system has a photoelectric conversion efficiency of 29.6% as well as a thermal efficiency of 18.52%. Parametric analysis results show that the thermal efficiency of the PVT system can be improved by increasing the inlet nano-fluids flow velocity as well as the ambient temperature, or by reducing the inlet nano-fluids temperature or convection heat transfer coefficient.

The current study is mainly focused on the monthly variation in the lightning flash rate (LFR) and related thermodynamic parameters using the data for the years 2000–2013, and the trend of lightning variation is explored. Lightning data are used from a lightning imaging sensor (LIS) and an optical transient detector (OTP) boarded on the tropical rainfall measuring mission (TRMM). Additionally, aerosol optical depth (AOD) data at 550 nm for the same period were considered from a Moderate Resolution Imaging Spectroradiometer (MODIS). The assessment of lightning and AOD using monthly data makes it difficult to study seasonal contributions, and higher-resolution (hourly) data may be more appropriate, but unfortunately, no data were available with a higher resolution than monthly. The dependency of LFR is also investigated using thermodynamic/dynamic parameters. The LFR shows a moderate correlation with a correlation coefficient of 0.56, 0.62, and 0.63 for AOD, CAPE, and vertical velocity, respectively. The increasing AOD in the pre-monsoon season is associated with higher lightning flash rates over this region. The possible sources of aerosols that cause an increase in lightning activities are identified from the classification of aerosols based on the characteristic values of the AOD and the Ångström exponent. The thermodynamic relation of the Product of Bowen ratio with the sum of the precipitation rate and evaporation rate has been used as a proxy to evaluate the lightning flash rate density over Srinagar, Uttarakhand region (78.55° E–79.05° E, 29.97° N–30.47° N), with nine models from the Coupled Model Inter-comparison Project-Phase 5 (CMIP5). The model-simulated LFR has also been used for the projection of lightning in the late 21st century, and the projected LFR over the study area shows a 7.41% increase during the (2079–2088) period as compared to the historic period (1996–2005). The results of the study region indicate caution in using any single climate variable as a proxy for projecting a change in the lightning–climate relationships in the scenario of global warming.

Leaves start to develop at the peripheral zone of the shoot apical meristem. Thereafter, symmetric and flattened leaf laminae are formed. These events are simultaneously regulated by auxin, transcription factors, and epigenetic regulatory factors. However, the relationships among these factors are not well known. In this study, we conducted protein-protein interaction assays to show that our previously reported Leaf and Flower Related (LFR) physically interacted with SWI3B, a component of the ATP-dependent chromatin remodeling SWI/SNF complex in Arabidopsis. The results of truncated analysis and transgenic complementation showed that the N-terminal domain (25–60 amino acids) of LFR was necessary for its interaction with SWI3B and was crucial for LFR functions in Arabidopsis leaf development. Genetic results showed that the artificial microRNA knockdown lines of SWI3B (SWI3B-amic) had a similar upward-curling leaf phenotype with that of LFR loss-of-function mutants. ChIP-qPCR assay was conducted to show that LFR and SWI3B co-targeted the promoters of YABBY1/FILAMENTOUS FLOWER (YAB1/FIL) and IAA carboxyl methyltransferase 1 (IAMT1), which were misexpressed in lfr and SWI3B-amic mutants. In addition, the association between LFR and the FIL and IAMT1 loci was partly hampered by the knockdown of SWI3B. These data suggest that LFR interacts with the chromatin-remodeling complex component, SWI3B, and influences the transcriptional expression of the important transcription factor, FIL, and the auxin metabolism enzyme, IAMT1, in flattened leaf lamina development.

Leaves start to develop at the peripheral zone of the shoot apical meristem. Thereafter, symmetric and flattened leaf laminae are formed. These events are simultaneously regulated by auxin, transcription factors, and epigenetic regulatory factors. However, the relationships among these factors are not well known. In this study, we conducted protein-protein interaction assays to show that our previously reported Leaf and Flower Related (LFR) physically interacted with SWI3B, a component of the ATP-dependent chromatin remodeling SWI/SNF complex in Arabidopsis. The results of truncated analysis and transgenic complementation showed that the N-terminal domain (25–60 amino acids) of LFR was necessary for its interaction with SWI3B and was crucial for LFR functions in Arabidopsis leaf development. Genetic results showed that the artificial microRNA knockdown lines of SWI3B (SWI3B-amic) had a similar upward-curling leaf phenotype with that of LFR loss-of-function mutants. ChIP-qPCR assay was conducted to show that LFR and SWI3B co-targeted the promoters of YABBY1 / FILAMENTOUS FLOWER (YAB1 / FIL) and IAA carboxyl methyltransferase 1 (IAMT1), which were misexpressed in lfr and SWI3B-amic mutants. In addition, the association between LFR and the FIL and IAMT1 loci was partly hampered by the knockdown of SWI3B. These data suggest that LFR interacts with the chromatin-remodeling complex component, SWI3B, and influences the transcriptional expression of the important transcription factor, FIL , and the auxin metabolism enzyme, IAMT1 , in flattened leaf lamina development. [ABSTRACT FROM AUTHOR]

• Diffusion couples performed up to 1200 °C for durations up to 100 h on UN and UO 2 with Fecral and Ti-modified stainless steel. • Chemical interaction between all combinations of materials was minimal and did not involve uranium. • No diffusion gradients were observed. • Ti in the modified stainless steel was found to form a discrete reaction phase with UN and UO 2. • Al in Fecral was found to form a discrete reaction phase with UN and UO 2. This study elucidates on the compatibility of fuel and clad materials for potential use in lead cooled fast reactors. Diffusion couples were performed on representative, unirradiated fuel forms (UN and UO 2) with potential cladding materials, a Ti-modified stainless steel (D9) and a FeCrAl alloy (Kanthal APMT). Temperatures of 800–1200°C, with durations up to 100 h, were investigated. Chemical interaction between all combinations of the fuel and cladding candidates was found to be minimal, with no diffusion gradients observed or extensive interface phases formed. When interaction occurred, it was in the production of a minor discrete phase. At the highest temperatures, D9 with UN resulted in the formation of a Ti-C-N phase. D9 with UO 2 resulted in the formation of a Ti-C-O phase. FeCrAl with UN resulted in the formation of an Al-N phase. FeCrAl with UO 2 results in the formation of an Al-O phase. Despite the observed mechanical bonding between fuel and cladding material, no evidence of any chemical interaction with the uranium in the fuel material was found. At the higher temperatures investigated (>1000°C) there was evidence of clad softening resulting in transfer of material to the fuel surface. [ABSTRACT FROM AUTHOR]

The highest efficiency in the usage of nuclear energy resources can be implemented in fast breeder reactors of generation IV. It is achieved thanks to the ability of consuming minor actinides (MAs) in energy production. One of the options to use this benefit is full recycling of MAs to close the nuclear fuel cycle. Monte Carlo burn up (MCB), an integrated burn-up calculation code, deals with the complexity of the burn-up process which is applied to the European Lead-cooled Fast Reactor (ELFR). MCB uses continuous energy representation of cross section and spatial effects of full core reactor model; however, it automatically calculates nuclide production in all possible reactions or decay channels. Multi-recycling of MAs can cause an intensified build-up of curium, berkelium and californium. Some of their isotopes are strong neutron emitters from spontaneous fission, which hinders fuel recycling. The implementation of a novel methodology for trajectory period folding allows us to trace the life cycle of crucial MAs from the beginning of the reactor life towards the state of adiabatic equilibrium. The result of the analysis performed is presented, showing the sources of strong contribution to the neutron production rate. The parametric sensitivity analysis method for selected nuclide reactions is applied, revealing sensitivity of transmutation chains for the production of neutron emitter isotopes.

Summary: The fuel rods in the lead‐based fast reactor (LFR) are usually supported by wire spacer, and the presence of the wire results in the complexity of the thermal‐hydraulic analysis in rod bundles. In this study, several previous experimental research about lead‐bismuth eutectic flowing through the rod bundle are investigated. Based on the collected experimental data, a series of flow and heat transfer correlations in the wire‐fixed rod bundle are compared and evaluated. The recommended correlations are then implemented into the in‐house sub‐channel analysis code SACOS‐PB. Furthermore, to account for the changes in the channel cross‐section parameters caused by the wire spacer, two optional wire geometric treatments were added to the code: axially‐averaged approach and axially‐varied approach with extra cross‐flow enhancement model. 19‐rod bundle and 61‐rod bundle tests are used to validate the optimized code. The deviations for the temperature differences of cladding and coolant in internal sub‐channels are within 25%, and the axially‐varied approach shows a higher prediction accuracy, which proves that the improvement of the SACOS‐PB are reliable. This work could provide a reference for the subsequent design and development of the LFR core. [ABSTRACT FROM AUTHOR]

Summary: Based on the favorable neutronic performance, T/H (thermal‐hydraulic) characteristics and inherent safety of the lead‐cooled fast reactor (LFR), University of Science and Technology of China performed a preliminary design of the 1000Mth medium‐power modular lead‐cooled fast reactor M2LFR‐1000. To solve the core optimization problem of M2LFR‐1000, an automatic core design code named CODOPT (core design and optimization) was developed, considering geometry, neutronics and thermal‐hydraulics. CODOPT starts from setting a number of the core design schemes, then evaluates their key neutronic and T/H characteristics automatically and finally finds the design schemes satisfying the design constraints. Using this code, the sensitivity analysis of the key design parameters (fuel rod diameter, pitch diameter ratio, and inlet coolant velocity) is carried out. Under the neutronic and T/H design constraints, the optimization problem of the core has been studied using a multi‐objective function, which tries to find an optimal design with larger Keff, flatter power distribution, and minimum temperature of cladding and pellet. Current results show that the overall neutronic and T/H performance of M2LFR‐1000 can be improved by choosing the design parameters presented in this paper. The automatic core design ideas adopted currently can also provide some reference for future LFR designs. [ABSTRACT FROM AUTHOR]