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Copyright of Journal of Xi'an Jiaotong University is the property of Editorial Office of Journal of Xi'an Jiaotong University and its content may not be copied or emailed to multiple sites or posted to a listserv without the copyright holder's express written permission. However, users may print, download, or email articles for individual use. This abstract may be abridged. No warranty is given about the accuracy of the copy. Users should refer to the original published version of the material for the full abstract. (Copyright applies to all Abstracts.)

Lead–bismuth eutectic (Pb–Bi) alloy, a liquid heavy metal, has garnered substantial attention as a candidate coolant for next-generation lead-cooled fast reactors (Gen-IV LFRs) and accelerator-driven systems (ADSs) owing to its exceptional nuclear and thermophysical properties. The demanding operational context within which these systems function poses significant challenges to the chemical and mechanical stability of traditional structural materials when in contact with lead–bismuth eutectic (LBE). Addressing the degradation and failure mechanisms of structural and fuel cladding materials in the presence of LBE is of paramount importance for the advancement and application of LFR and ADS technologies. Among various candidate materials, ferrite/martensite (F/M) steel has been considered an ideal candidate for LBE-cooled reactor fuel cladding owing to its excellent mechanical properties, high-temperature mechanical performance, radiation resistance, and lower coefficient of thermal expansion. This study focuses on P91 F/M steel. Dynamic tests were performed for up to 3000 h in uncontrolled oxygen LBE at a relative flow velocity of 5 m·s−1 and temperatures of 350 ℃ and 450 ℃. The surface and cross-sectional corrosion morphologies of the alloy under different temperatures and exposure times were systematically analyzed by scanning electron microscopy. The composition and structural evolution of the oxide layer formed on P91 steel were summarized, and the mechanism of oxide layer spallation was proposed. In addition, electron backscatter diffraction was used to analyze the grain size, stress distribution, and proportion of large-angle grain boundaries in different impact angle regions between LBE and the samples, providing a detailed discussion of the erosion mechanism of the alloy in these regions. At 350 ℃, the oxide layer is a multilayer structure comprising a porous Fe3O4 layer, Fe-Cr spinel layer, and inner oxidation zone undergoing a dynamic equilibrium process of “formation–spallation–reformation”. At 450 ℃, the oxidation-corrosion phenomenon is severe, with LBE penetration observed in addition to oxidation. The corrosion characteristics on the sample surface vary significantly across different impact angle regions. The severity of surface damage is ranked as follows: 30° > 90° > −90°. The oxide layer in the 30° angle region completely spalls, with LBE penetrating into the matrix. The 90° angle region shows spalling of the porous outer oxide layer, leaving only the inner oxide layer, with the inner matrix uneroded by LBE. The −90° angle region maintains an intact oxide layer structure that is free from LBE erosion. In this work, the corrosion–erosion behaviors of P91 steel in high-velocity (5 m·s−1) LBE are studied, the formation and spallation mechanisms of the oxide scales are elucidated, and the erosion damage mechanism of the alloy are revealed, providing experimental data and references for the development of structural or cladding materials for China’s fourth-generation nuclear reactors and the study of their corrosion mechanisms in LBE.

Traditional text-based system engineering, which has been used in the design and application of passive residual heat removal system (PRHRS) for lead-cooled fast reactors, is prone to several problems such as low development efficiency, long iteration cycles, and model ambiguity. This study aims to effectively address the above-mentioned problems by adopting a model-based system engineering (MBSE) method, which has been preliminarily applied to meet the design requirements of a PRHRS. The design process has been implemented based on the preliminary design of the system architecture and consists of three stages: top-level requirement analysis, functional requirements analysis, and design requirements synthesis. The results of the top-level requirements analysis and the corresponding use case diagram can determine the requirements, top-level use cases, and scenario flow of the system. During the functional requirements analysis, the sequence, activity, and state machine diagrams are used to develop the system function model and provide early confirmation. By comparing these sequence diagrams, the requirements for omissions and inconsistencies can be effectively checked. In the design requirements synthesis stage, the Analytic Hierarchy Process is used to conduct preliminary trade-off calculations on the system architecture, after which a white box model is established during the system architecture design. Through these two steps, the analysis and design of the system architecture are ultimately achieved. The resulting system architecture ensures the consistency of the design requirements. Ultimately, a functional hazard analysis was conducted for a specific incident to validate case requirements and further refine the system architecture. Future research can further reduce the design risk, improve the design efficiency, and provide a practical reference for the design and optimization of PRHRS in digital lead-cooled fast reactors.

Steam generator tube rupture (SGTR) accident is one of the accident scenarios that must be considered during the design and safety analysis process of lead-cooled fast reactors (LFR). Focused on this topic, numerous experiments and theoretical research have been conducted in recent decades, which could be generally categorized into two aspects: on the one hand, separation effect experiments and theoretical analysis focusing on the specific phenomena of lead-water interaction process; on the other hand, engineering test facilities and numerical simulations focusing on the overall process of LFR SGTR accident, mainly concentrated on the evolution of macro parameters such as system pressure and lead pool temperature during SGTR accident. In this paper, firstly, the basic mathematical and physical models of ACENA code was introduced; then the code was validated against the LBE-N2 two-phase flow experiment HESTIA-2, the KYILN-Ⅱ-S LBE-H2O interaction experiment and the analytical solution of the point-kinetic neutronic equations; after that the Eulerian multi-phase analysis code ACENA was applied to simulate the postulated unprotected SGTR process of European pool-type lead-cooled fast reactor ALFRED. Attention was paid to the simulated migration process of steam bubbles in the lead pool as well as the evolution of safety parameters such as maximum cladding temperature, fuel centerline temperature, relative nuclear power and reactor vessel pressure. Furthermore, the sensitivity of ruptured tube quantities, lead coolant circulation paths and neutron dynamics parameters on calculation results were analyzed. This paper’s study demonstrates that the Ishii-Chawla-Suzuki interfacial drag coefficient model could accurately predict the diffusion and migration characteristics of rising bubbles in circular/annular LBE flow channels when combined with the interfacial area concentration transport model proposed by Ishii et al; The simulation of KYLIN-Ⅱ-S experiment shows that ACENA code could reasonably reproduce phenomena such as system pressure fluctuations and temperature transients of liquid lead during the interaction process between LBE and water; the calculated results of ALFRED reactor’s key thermal-hydraulic parameters under hot full power condition by ACENA code are generally consistent with results obtained by internationally recognized one-dimensional system code TRACE/FRED, proving the reliability of ACENA code when applied to full reactor system level; The calculation of the hypothetical SGTR accident of the ALFRED reactor verifies the ACENA code’s capability of simulating the complex multi-component multi-phase flow, and the calculation results reveal that designing the primary coolant flow path rationally and reducing the number of ruptured pipes are of great significance to mitigating the potential consequences of LFR SGTR accidents. This work can provide reference for the safety analysis work of China’s pool-type lead-cooled fast reactors.

Lead-cooled fast reactors (LFRs) have a wide range of application scenarios, which require the thermal-hydraulic characteristics of LFRs to be reliable. In the present paper, the Lead-cooled fast reactor Thermal-Hydraulic Analysis Code LETHAC was developed, including the models of pipe, heat exchanger, and pool. To verify the correctness of LETHAC, two experimental facilities and three experimental cases were selected, including GFT and PLOFA tests for NACIE-UP and Test-1 for CIRCE. The calculated results show the same and consistent trend with the experimental data, but there are some discrepancies. It can be found that LETHAC is suitable and reliable in predicting the transient behavior of lead-cooled system.

Copyright of Atomic Energy Science & Technology is the property of Editorial Board of Atomic Energy Science & Technology and its content may not be copied or emailed to multiple sites or posted to a listserv without the copyright holder's express written permission. However, users may print, download, or email articles for individual use. This abstract may be abridged. No warranty is given about the accuracy of the copy. Users should refer to the original published version of the material for the full abstract. (Copyright applies to all Abstracts.)

2D and 3D ultrasonic imaging has so far not been demonstrated in pure molten lead in the open literature. In this study the development of such an ultrasonic device for imaging is outlined and results from testing at 380 °C in lead are presented. The main difficulties were found to be achieving then maintaining suitable wetting while ensuring suitable durability of the device, both due to the harsh nature of molten lead and the elevated temperatures. The successful detection and imaging (2D and 3D), of differently shaped targets, where the features were above the size of the transmitted ultrasound beam was demonstrated.

The lead-cooled fast reactor (LFR) is one of the most promising Generation-IV nuclear designs currently under development in Europe, China, and the United States. LFRs can ensure enhanced performance and minimal waste production thanks to a closed fuel cycle, but they also have some issues that need to be addressed. One of the most critical is the long-term degradation process initiated in structural materials exposed to liquid Pb. The present state of the art has shown that commercial austenitic steels, such as American Iron and Steel Institute 316L and 15-15Ti can be adopted as structural materials in Pb environments up to 480°C, beyond which they start to experience the dissolution of constituting alloying elements (Ni, Cr, and Fe) if not protected by a coating or by surface modification. In more recent years, a lot of research effort has been done in order to develop new coating technologies and new base materials for operation with liquid Pb at higher temperatures. Among the newest alloys, alumina-forming austenitic (AFA) steels have gained interest in the research community because of their promising corrosion resistance results even at temperatures of 600°C. In this framework, an experimental campaign has been run at the Research Center ENEA of Brasimone that aims to characterize the behavior of two different AFA steels (with low and high Ni content in their composition) in static Pb at 650°C and 750°C with a moderate low oxygen concentration (10−6 wt %). After exposure, the AFA steels were characterized from the point of view of the morphology and composition, and the results are presented and discussed here. [ABSTRACT FROM AUTHOR]

In the framework of the ALFRED research and development program, the ATHENA facility will be constructed for thermal-hydraulic analysis of full-scale ALFRED components and systems. The source system of the facility is the core simulator, which aims to be representative of an ALFRED average fuel assembly. Computational fluid dynamics (CFD) codes are gaining attention for the analysis of complex systems in pool-type reactors since they are able to reproduce three-dimensional phenomena. In this paper, a multiscale approach based on porous media is proposed to reduce the computational cost of the core simulator CFD model. The multiscale approach starts with the detailed simulation of the infinite lattice domain of the fuel assembly to characterize the porous media hydraulic behavior. Then the porous media are applied in the system model. Three different approaches are investigated: (1) adopting a single porous media for the entire fuel assembly, (2) representing the bundle with two porous domains, and (3) adopting the so-called hybrid medium. The results have been compared with the reference detailed CFD simulation for performance evaluation. The first step of the analysis is the application of the multiscale approach on the CIRCE fuel pin simulator to carry out a turbulence model validation against experimental data and a comparison of the three approaches with a proven CFD model. Then the approach is applied on the ATHENA core simulator exploiting the CIRCE results. The results obtained with the porous media models are compared with a detailed CFD simulation of the core simulator to evaluate the performance of the three approaches. Eventually, the best solution is applied on a model of the entire ATHENA core simulator integrated with the feeding region. The model is tested also in transient conditions. The numerical experiment demonstrates the effectiveness of the multiscale approach in reducing the computational cost while maintaining high accuracy in representing the quantities of interest. [ABSTRACT FROM AUTHOR]

Westinghouse is developing a lead-cooled fast reactor (LFR) as its next-generation utility-scale nuclear power plant. To support its development, Westinghouse and its partners are building 10 test facilities to demonstrate key LFR phenomena, materials, and components in liquid lead. These test infrastructures are distributed across several institutions, and this paper focuses on those located within Westinghouse. It describes three state-of-the-art test rigs being installed in the Westinghouse facility in Springfields, United Kingdom, to test materials in liquid lead and to investigate key LFR phenomena, which will also be used to validate modeling and simulation tools. These facilities address material corrosion/erosion testing (MELECOR), lead freezing and under-lead viewing testing (LEFREEZ), and primary heat exchanger failure testing (LEWIN). This paper also describes the first liquid-lead test system to become operational at Westinghouse at the company's site in Churchill, Pennsylvania, i.e., a test rig named HELMET to conduct tensile tests in a molten-lead environment to investigate the potential susceptibility of LFR candidate materials to liquid-metal embrittlement. [ABSTRACT FROM AUTHOR]

The accurate calculation of reactor core heating is vital for the design and safety analysis of reactor physics. However, negative KERMA factors may be produced when processing and evaluating libraries of the nuclear data files ENDF/B-VII.1 and ENDF/B-VIII.0 with the NJOY2016 code, and the continuous-energy neutron cross-section library ENDF71x with MCNP also has the same problem. Negative KERMA factors may lead to an unreasonable reactor heating rate. Therefore, it is important to investigate the influence of negative KERMA factors on the calculation of the heating rate. It was also found that negative KERMA factors can be avoided with the CENDL-3.2 library for some nuclides. Many negative KERMA nuclides are found for structural materials; there are many non-fuel regions in fast reactors, and these negative KERMA factors may have a more important impact on the power distribution in non-fuel regions. In this study, the impact of negative KERMA factors on power calculation was analyzed by using the RBEC-M benchmark and replacing the neutron cross-section library containing negative KERMA factors with one containing normal KERMA factors that were generated based on CENDL-3.2. For the RBEC-M benchmark, the deviation in the maximum neutron heating rate between the negative KERMA library and the normal library was 6.46%, and this appeared in the reflector region. In the core region, negative KERMA factors had little influence on the heating rate, and the deviations in the heating rate in most assemblies were within 1% because the heating was mainly caused by fission. However, in the reflector zone, where gamma heating was dominant, the total heating rate varied on account of the gamma heating rate. Therefore, negative KERMA factors for neutrons have little influence on the calculation of fast reactor heating according to the RBEC-M benchmark. [ABSTRACT FROM AUTHOR]

The accurate calculation of reactor core heating is vital for the design and safety analysis of reactor physics. However, negative KERMA factors may be produced when processing and evaluating libraries of the nuclear data files ENDF/B-VII.1 and ENDF/B-VIII.0 with the NJOY2016 code, and the continuous-energy neutron cross-section library ENDF71x with MCNP also has the same problem. Negative KERMA factors may lead to an unreasonable reactor heating rate. Therefore, it is important to investigate the influence of negative KERMA factors on the calculation of the heating rate. It was also found that negative KERMA factors can be avoided with the CENDL-3.2 library for some nuclides. Many negative KERMA nuclides are found for structural materials; there are many non-fuel regions in fast reactors, and these negative KERMA factors may have a more important impact on the power distribution in non-fuel regions. In this study, the impact of negative KERMA factors on power calculation was analyzed by using the RBEC-M benchmark and replacing the neutron cross-section library containing negative KERMA factors with one containing normal KERMA factors that were generated based on CENDL-3.2. For the RBEC-M benchmark, the deviation in the maximum neutron heating rate between the negative KERMA library and the normal library was 6.46%, and this appeared in the reflector region. In the core region, negative KERMA factors had little influence on the heating rate, and the deviations in the heating rate in most assemblies were within 1% because the heating was mainly caused by fission. However, in the reflector zone, where gamma heating was dominant, the total heating rate varied on account of the gamma heating rate. Therefore, negative KERMA factors for neutrons have little influence on the calculation of fast reactor heating according to the RBEC-M benchmark.

The hydrodynamic and thermal interaction of water with the high-temperature melt of a heavy metal was studied via the Volume-of-Fluid (VOF) method formulated for three immiscible phases (liquid melt, water, and water vapor), with account for phase changes. The VOF method relies on a first-principle description of phase interactions, including drag, heat transfer, and water evaporation, in contrast to multifluid models relying on empirical correlations. The verification of the VOF model implemented in OpenFOAM software was performed by solving one- and two-dimensional reference problems. Water jet penetration into a melt pool was first calculated in two-dimensional problem formulation, and the results were compared with analytical models and empirical correlations available, with emphasis on the effects of jet velocity and diameter. Three-dimensional simulations were performed in geometry, corresponding to known experiments performed in a narrow planar vessel with a semi-circular bottom. The VOF results obtained for water jet impact on molten heavy metal (lead–bismuth eutectic alloy at the temperature 820 K) are here presented for a water temperature of 298 K, jet diameter 6 mm, and jet velocity 6.2 m/s. Development of a cavity filled with a three-phase melt–water–vapor mixture is revealed, including its propagation down to the vessel bottom, with lateral displacement of melt, and subsequent detachment from the bottom due to gravitational settling of melt. The best agreement of predicted cavity depth, velocity, and aspect ratio with experiments (within 10%) was achieved at the stage of downward cavity propagation; at the later stages, the differences increased to about 30%. Adequacy of the numerical mesh containing about 5.6 million cells was demonstrated by comparing the penetration dynamics obtained on a sequence of meshes with the cell size ranging from 180 to 350 µm. [ABSTRACT FROM AUTHOR]

Renewable technologies using solar input have varied electrical power production during periods of low solar irradiance caused by cloud coverage, seasonal changes, and time of day. Nuclear power plants can load follow, but due to low operating costs and high fixed costs, this capability is often not financially appealing. Implementing thermal energy storage (TES) within a synergistic solar and nuclear power cycle allows for storage during low demand periods and increased power production during high demand periods, effectively increasing dispatchability. In this paper, we examine the thermodynamic performance of an integrated system that includes concentrating solar power (CSP) and a lead-cooled fast reactor (LFR). These technologies are selected due to their similar operating temperatures, allowing for utilization of established TES technologies. Both the CSP and LFR system produce thermal power that is sent to a supercritical steam-Rankine cycle (SSRC). The SSRC model is designed to be implemented into a larger integrated energy system (IES) which contains multiple communicating models. The IES generates CSP and LFR heat profiles then utilizes the SSRC to output calculated metrics including power generation, heat rejection, and cycle efficiency.

BackgroundTraditional text-based system engineering in the design and application of passive residual heat removal systems (PRHRS) for lead-cooled fast reactors has several requirement problems, such as low development efficiency, a long iteration cycle, and model ambiguity.PurposeThis study aims to effectively address the aforementioned problems encountered in the PRHRS of lead-cooled fast reactors by adopting a model-based system engineering method.MethodsThe model-based system engineering (MBSE) method was preliminarily applied to the design requirement analysis of a PRHRS for lead-cooled fast reactors. During the design requirements study, the design process was combined with the preliminary design of the system architecture, comprising three parts: requirements analysis, functional analysis, and design synthesis.ResultsThe generated requirement and use case diagrams describe the system requirements and determine the top-level use cases of the system in the requirement analysis stage. The time sequence, activity, and state machine diagrams form the system function model and provide early confirmation and verification in the functional analysis stage. Finally, the white box model realizes the analysis and design of the system architecture in the design synthesis stage.ConclusionsThe system architecture designed by this method ensures the consistency of the design requirements. In the future, it will further reduce the design risk, improve the design efficiency, and provide an application reference for the design and optimization of digital lead-cooled fast reactors' passive residual heat removal systems.

Lead-cooled fast reactor (LFR) is an important reactor type of the Generation IV nuclear energy, and the main coolant pump (MCP) is the only and most critical power equipment in primary circuit of LFR. The shutdown idling performance of the MCP, is one of the most important safety indicators in the event of a station blackout accident. When the motor loses its power supply, the MCP needs to run under the inertia of the flywheel to ensure the coolant flow for a short time. This study mainly focuses on the idling mathematical model of the MCP under power failure accidents. In this study, taking the shutdown idling performance of the MCP as the optimization goal, the parametric optimization analysis of the impeller and diffuser structure of the MCP was carried out. Specifically, based on the platform ISIGHT, the software of CFturbo, TurboGrid, CFX, Matlab and FLOWMASTER are systematically integrated, the automated 3D modeling, automatic meshing and automatic CFD calculation of different hydraulic model of the MCP are realized. And the neural network mathematical model between the geometric parameters of MCP and the idling performance indicators is established. The study shows that there is a strong linear relationship between the idling performance of the MCP and its head and shaft power, and meridian parameters of Δβ 2 , Z 1 , Φ 1h and the radial surface parameters of Δβ 1 and Δβ 2 , Z 1 and Φ 1h have the greatest impact on the idling performance of MCP. [ABSTRACT FROM AUTHOR]

Potentiometric oxygen sensors play a crucial role in measuring dissolved oxygen concentration in lead-cooled fast reactors (LFRs). The choice of the solid electrolyte material, including yttria partially stabilized zirconia (YPSZ) and yttria stabilized zirconia (YSZ), exerts a significant influence on the sensors' performance characteristics. In this study, the Potentiometric oxygen sensors with YPSZ and YSZ solid electrolyte were tested within the oxygen-saturated lead-bismuth eutectic (LBE) at the temperature range from 473 K to 773 K. The ionic conductivities of the solid electrolytes were obtained by alternating current impedance. The crystal phase changes of YSZ and YPSZ were compared by variable temperature XRD. Notably, the oxygen ion channels begin to form in solid electrolytes at 523 K, with full activation occurring at 643 K for YSZ and 723 K for YPSZ. The electromotive force (EMF) of YSZ and YPSZ-sensors possess good accuracy and stability between 623 and 773 K. Below 623 K, the EMF of YPSZ-sensor is consistently deviated from the theoretical value. These variations in activation temperatures are attributed to differences in Y3+ content, resulting in distinct operational temperature ranges for electrolytes. [ABSTRACT FROM AUTHOR]

• Core fuel multitype PD ratio layout scheme of the Offshore Fixed Lead-Cooled Reactor are proposed and effectively reduces the radial power peak factor at mid- and end-of-life, changes the trend of radial power peak factor with burnup and reduces the reactivity fluctuation. • The MA nuclide doping study based on core fuel multitype PD ratio layout scheme has been carried out. • The analysis of preliminary physical design of the Offshore Fixed Lead-Cooled Reactor and the evaluation of the modeling equivalent method based on SARAX has been carried out. The Offshore Fixed Lead-Cooled Reactor is a Small Long-Life Lead-Cooled Fast Reactor. Key physical parameters of the initial core design scheme were calculated to evaluate the accuracy of the modeling equivalent method and verify the rationality of the core physical design scheme. This study compared the results of the deterministic code SARAX with those of the Monte Carlo code OpenMC. Key physical parameters were calculated and found to be consistent, which demonstrating the suitability of the SARAX on design and analyses for the Offshore Fixed Lead-Cooled Reactor. The preliminary physical design has been confirmed to meet the design requirement. To address the issue of excessive radial power peak factor at the end of life, a core fuel multitype PD ratio layout scheme was designed, and studies on Minor Actinides nuclide doping were carried out. The radial power peak distribution factor decreased at mid- and end-of-life, and the reactivity fluctuation throughout lifetime was reduced. [ABSTRACT FROM AUTHOR]

A combined experimental and modelling study into the Pb-Mo-O system has been conducted in view of the safety analysis for lead-cooled nuclear systems. The thermal expansion and low-temperature heat capacity of the ternary compounds PbMoO 4 and Pb 2 MoO 5 have been determined experimentally, as well as the melting enthalpy of PbMoO 4. Moreover, XANES measurements have confirmed the hexavalent oxidation state of Mo in PbMoO 4 and Pb 2 MoO 5. A thermodynamic model of the ternary system including the ternary phases PbMoO 4 , Pb 2 MoO 5 and Pb 5 MoO 8 has also been developed in this work based on the CALPHAD methodology. For the first time, an ionic two-sublattice model is used for the liquid phase, while the compound energy formalism is used for the solid phases. [Display omitted] • A thermodynamic model of the Pb-Mo-O system is described. • The low-temperature heat capacity and thermal expansion of Pb 2 MoO 5 and PbMoO 4 have been measured. • The melting enthalpy of PbMoO 4 has been determined. • XANES of Pb 2 MoO 5 and PbMoO 4 at the Mo K-edge are reported. [ABSTRACT FROM AUTHOR]

Generation IV (Gen-IV) nuclear reactor designs are receiving increasing attention because of their potential to achieve key goals in terms of sustainability, safety, reliability, and economics. One of the six technologies selected in the Gen-IV program is the lead-cooled fast reactor (LFR). This work focuses on LFRs, examining both pure lead and lead-bismuth eutectic designs. Through two systematic literature reviews, we consolidate the state-of-the-art in economics and finance for LFRs. The first review considers scientific literature and retrieves 12 articles. The second focuses instead on industrial literature, resulting in 12 additional documents. Economics literature is very scarce and sometimes of low quality. Economic estimations for the specific capital cost [$/kWe] and the cost of electricity [$/MWh] can vary by an order of magnitude (1500–25000 $/kWe and 30–350 $/MWh, respectively), while design organizations typically do not publicly share financial details. Finance literature is almost nonexistent. We report, in the final part of the work, notable knowledge gaps and further possible research areas. • Comprehensive economic evaluations for LFRs are missing. • Economic estimations for LFRs vary by an order of magnitude. • Reported LFR cost ranges are 1500–25000 $/kWe and 30–350 $/MWh. • Financial information of LFRs is rarely disclosed to the public. • A viable financial model description is missing for LFRs. [ABSTRACT FROM AUTHOR]

Lead-cooled fast reactor (LFR) is known as a typical reliable new type reactor by the US Department of Energy (DOE), and the potential accidents of LFR are necessary to assess. Steam generator tube rupture (SGTR) accident can cause high-pressure hyperbaric steam-water jetting into liquid lead bismuth eutectic (LBE) pool, and leads to a rapid pressure rising in the reactor primary system together with a potential pressure wave crashing the structures. This paper focuses on the computational fluid dynamics (CFD) analyses of SGTR experimental research on the LIFUS5/MOD2 experiment facility. A two-dimensional symmetrical slice with 1 million 570 thousand structured mesh was established for the experimental reaction vessel, and each side except the injection nozzle were set as adiabatic no slip boundary in order to achieve the non-decompression process. In the aspect of multiphase model settings, the Eulerian-Eulerian two-phase flow model was adopted to track the steam-LBE interface in the transient simulation, and the CSF surface tension polynomial together with Lee substituted evaporation model were considered. In addition, the injection of high-pressure water in each time step was separated to two parts: latent heat and injected steam, so as to calculate both endothermic and jet of flash evaporation simultaneously. Among them, the latent heat was realized by user-defined functions (UDF), and the injected steam was adopted as compressible real gas with R-K gas state equation. Based on the above settings, three established thermal-hydraulics phenomena including pressure rising of LBE environment together with pressure wave spreading, level fluctuation of liquid environment with argon entrainment and level rising of liquid environment and diffusion of injected steam were analyzed. From the results a conclusion can be reached that CFD numerical simulation has little error on calculating pressure rising with pressure wave spreading. In addition, the peak value of pressure wave rises with back pressure of water increasing, and the pressure of partial steam cavity is lower than that of LBE pool around it, in turn inhibits the occurrence of its explosion. Meanwhile, LBE in the middle firstly washes up and impacts the structures, and then entrains the argon when it falls back, in turn causes the combination of argon and steam cavity, or the secondary segmentation to small argon bubbles. Moreover, the diffusion of the steam cavity along the radial direction becomes more intense with the increase of the axial height, accompanies by the steam cavity divided from the center to the edge along the radius.

During scram, the Control Rod Assembly (CRA) is quickly dropped into the core and as well, if any of the operating limits are exceeded, the CRA is dropped into the core within a stipulated time to shut down the reactor power as soon as possible. In this study, the Computational Fluid Dynamics (CFD) approach was used to investigate the CRA drop dynamics of a lead-based research reactor. To simulate the flow field around the CRA in the guide tube, a 3-dimensional model of the CRA in the LBE-filled guide tube was developed and discretized; and the averaged Navier-Stokes equations coupled with the dynamic mesh method were adopted. Considering the large mesh deformation in the LBE coolant domain while the CRA drops, the recently developed FSI method in the CFX code, namely the rigid body approach, was adopted, which falls under the monolithic method. In this method, the translational CRA wall, which is partially immersed in the LBE, was set as a rigid body. It has the advantage of updating and improving the mesh quality through the mesh and re-meshing technique during the process of computation. Compared with the results of the work done in the available literature, the CFD model proved to be applicable and reliable. From the results, the inherent high density among the LBE flow characteristics had the most influence on the drop time. The mass of the CRA impacts its driving force so that the drop time reduces when the CRA mass is increased. In conclusion, the method used in this study can be applied to compute and predict significant parameters which can serve as a reference for a suitable design of the CRA and its drive mechanism in the case of modification for safety.

• The main coolant pump (MCP) was studied by the combination of experiment and numerical simulation. • The accuracy of the numerical method is verified by the measured data of MCP head and efficiency. • The distribution of velocity, pressure and turbulent kinetic energy inside MCP were calculated by numerical simulation. • The increase of coolant energy mainly depends on the joint action of the second half of the blade and the shroud side. As the key component of the primary circuit of a lead-cooled fast reactor (LFR), the hydraulic performance and the internal flow characteristics of the main coolant pump (MCP) are significantly important to the long-term safe and efficient operation of LFR. The present study reports an experimental and numerical investigation of the hydraulic performance and the internal flow characteristics of a centrifugal pump, which is considered one of the candidates for the MCP in LFR. Based on the experimental platform of Liquid Metal-Supercritical Fluid coupled heat transfer (LMSF) of Xi'an Jiaotong University, the hydraulic performances of a centrifugal pump like the pump head and its efficiency are studied and the measured data is also used to validate the accuracy of the numerical method which are employed to study the flow characteristics of MCP, as velocity, pressure, turbulent kinetic energy (TKE). The results show that the deviation of experimental and numerical simulation results is less than 5%. The high velocity region in the impeller channel is concentrated near the hub side of the blade inlet, and the significant pressure difference of the blade surface is primarily concentrated in the second half and shroud region. The inlet area of the blade exhibits a concentration of high TKE and the circular distribution of TKE within the impeller exhibits a periodic pattern, with its frequency being consistent with the number of blades. [ABSTRACT FROM AUTHOR]

• A detailed flow field analysis of the multi-pitch wire-wrapped fuel rod case was carried out using large eddy simulation method. • A unidirectional fluid–structure coupling method was used to simulate the vibration of the fuel rod due to the fluctuation of fluid excitation force. • A combination of time domain and frequency domain analysis was used to analyze the fluid excitation force fluctuations and fuel rod vibration. • The effects of different fluid flow rates, different fixation methods of fuel rods, different fluid physical parameters and different winding pitches on flow-induced vibration under axial flow are analyzed. Fixing fuel rods with wires is a common positioning method in reactors, especially fast reactor cores. Due to the presence of wires, unbalanced or even fluctuating excitation forces are generated when the fluid flows through the fuel rods, resulting in fuel rod vibration. In this paper, the effects of different wire pitches, different inlet velocities and different fuel rod fixation methods on wire-wrapped fuel rod vibration are analyzed using the FSI numerical method. The results show that changing the wire pitch, inlet velocity and fuel rod fixation method can change the magnitude and fluctuation of the fluid pressure on the surface of the fuel rod, which can further change the results of the vibration response of the fuel rod. However, changing any of these factors does not cause the fuel rod to resonate. [ABSTRACT FROM AUTHOR]

The lead-cooled fast reactor (LFR) is one of the most promising fast neutron reactors using molten lead or the lead–bismuth eutectic (LBE) alloy as a coolant. Under postulated severe accidents, the fuel rod of LFR may be damaged, which would cause the release of fission gas, and the migration of fission gas bubbles in the reactor molten pool will affect the release and absorption of radioactive substances in the reactor. In this paper, a three-dimensional numerical study on the release and migration behavior of fission gas in the molten LBE pool of LFR is carried out based on the volume of fluid method. The bubbles are continuously released by gas injection, and the research mainly focuses on the detachment time, the rising velocity, and the size of the bubble when it detaches at the orifice. The coalescence of bubbles is observed, and the acceleration effect of the bubble wake is confirmed. The distribution of the bubble terminal rising velocity with diameter has no simple or linear relationship. The effects of the gas injection velocity, the release depth, and the gas injection angle are studied. A lower gas injection velocity will delay the detachment and reduce the size of the bubble. The increase of release depth tends to release smaller bubbles. The bubbles released from a vertical surface will attach to the wall. The simulations and theoretical analysis are comparable and have similar tendencies. The distribution of the bubble terminal rising velocity with equivalent diameter may predict the migration behavior of bubbles in molten LBE.

To study the model predictive control (MPC) for a lead-cooled fast reactor's core power control system (LFR), firstly, the LFR core is established with the state space model. Then, to establish an LFR core power control system, a predictive model controller is used. Finally, the conditions of 20pcm step reactivity and 5% step down of coolant inlet temperature are introduced to study the control characteristic. The maximum overshot (MO) and the transient time are calculated in the time domain, and the stability of the core power control system is analyzed using the Nyquist and Bode diagrams in frequency domains. The result shows that the MPC controller and PID controller are feasible in core power control, and the core power control systems are closed-loop stable systems.

The lead-cooled fast reactor (LFR) engenders novel requisites for the energy conversion system when applied to maritime ships and distributed energy supply systems. The supercritical carbon dioxide (sCO 2) Brayton cycle, with the advantages of high efficiency, space-saving, simplicity, and flexibility, is a promising candidate for LFR. Here, we present the first study on the off-design operation of the sCO 2 cycle for LFR, with the turbine and compressor arranged on a single shaft (coaxial-shaft) or two shafts (split-shaft). The off-design performances of the system are investigated using four rotational speed (RS) control-based hybrid control strategies. We showed that the split-shaft system controlled by separate turbine speed and compressor speed has a wider operating space of power load than the coaxial-shaft system. Among the four investigated control strategies, the RS-inventory hybrid control can achieve the best thermal efficiency during load variation by operating at the turbine speed and mass flow rate corresponding to the optimal turbine isentropic efficiency. The maximum operating space of power load (0%–100 %) can be achieved by the RS-turbine inlet pressure hybrid control and the RS-bypass hybrid control. Our study can provide guidance for the flexible and safe part-load operation for small-scale modular Gen-IV nuclear reactors. • Off-design performance of sCO 2 cycle for LFR is evaluated. • Arrangements of turbomachines in coaxial- and split-shaft are compared. • Four hybrid control strategies based on the rotational speed control are studied. • Operating space of thermal efficiency and power load are analyzed. [ABSTRACT FROM AUTHOR]

• Two sets of Offshore Fixed Lead-Cooled Reactor core optimization schemes are proposed and demonstrated the effectiveness of fuel zoning in reducing end-of-life radial power peak factor. • The application in long-life small lead reactors of SARAX and the core equivalent homogenization model have been verified. • The analysis of initial physical design of Offshore Fixed Lead-Cooled Reactor, proved the meeting of design requirement and found the problem of high radial power peak factor at end-of-life. The Offshore Fixed Lead-Cooled Reactor is a Small Long Life Lead Cooled Fast Reactor. In order to verify the rationality of the physical design and applicability of SARAX and the core equivalent model method to the corresponding reactor type, the calculation of the key physical parameters of the Offshore Fixed Lead-Cooled Reactor and the physical design characteristics of the core were investigated and compared with the Monte-Carlo code OpenMC. The initial physical design has been shown to meet the design requirement. The reactivity coefficient of the core during life were calculated and verified, which proved the rationality of the initial core design. In order to solve the problem of high radial power peak factor at end-of-life, analysis of radial fuel zoning was conducted and the core optimization was carried out. The calculation of the key physical parameters of the optimization schemes was conducted, which proved the rationality of optimization schemes. [ABSTRACT FROM AUTHOR]

The goal of this work is to develop a high-order numerical technique to solve a two-dimensional fractional neutron diffusion (FND) model describing dynamical behavior of a lead-cooled fast reactor. This method is based on L 1 technique for temporal discretization and a high-order compact Alternating Direction Implicit (ADI) finite difference scheme for the spatial discretization. Numerical simulations are conducted to demonstrate the applicability of the method. We calculate the neutronic flux in the core and examine how the order of the temporal-fractional derivative influences the neutronic flux. The results obtained with FND model for different values of anomalous diffusion coefficient are compared against those obtained with P1 model and classical neutron diffusion equation. In addition, the computational time (CPU time) is provided in order to justify the computational efficiency of the proposed method. • Numerical method for two-dimensional fractional neutron diffusion model. • Anomalous diffusion (non-Fickian) in lead-cooled fast reactors. • ALFRED nuclear reactor dynamic analysis for different coefficient of anomalous diffusion. • Comparisons with classical neutron diffusion theory (normal diffusion) and P1 equations. [ABSTRACT FROM AUTHOR]

We have investigated the dissolution and retention of several metallic fission products and iodine in lead coolant under conditions relevant to accident scenarios in lead-cooled fast reactors (LFRs). First, the inventory of these fission products was calculated under some possible LFR operating conditions using a numerical model at various reactor operation times ranging from 6 months to 2 years. Secondly, low concentrations (1 to 2 wt%) of these fission products were dissolved in molten lead and the measured solubilities compared to the calculated reactor inventory, with the exception of iodine and niobium, which volatilized and did not digest, respectively. Finally, the transient dissolution behavior of certain selected fission products was examined and shown to take place on the timescale of hours. The findings of this study have bearings on the modeling of LFR accident scenarios and the overall safety analysis of this class of reactors. In addition, the results suggest that the behavior of molybdenum in molten lead at temperatures in excess of 800 °C is more complex than has previously been supposed and merits future study. • Inventory of 9 major fission products calculated for LFRs. • Solubility of these fission products in molten Pb is investigated. • Pb retains entire inventory for most fission products under accident conditions. • Transient experiments suggest dissolution proceeds on the order of hours. [ABSTRACT FROM AUTHOR]

Variable renewable energy availability has increased the volatility in energy prices in most markets. Nuclear power plants, with a large ratio of capital to variable costs, have historically operated as base load energy suppliers but the need for more flexible operation is increasing. We simulate the techno-economic performance of a 950 MW t nuclear power plant, based on the Westinghouse lead-cooled fast reactor, coupled with molten salt thermal storage as a method for flexible energy dispatch. We use the System Advisor Model to model the nuclear reactor thermal power input and power cycle operating modes. We combine this robust engineering model with a mixed-integer linear program model for optimized dispatch scheduling. We then simulate the coupled nuclear and thermal storage system under different market scenarios with varying price volatility. We find that the coupled plant outperforms the base plant under markets where energy price peaks fluctuate by a factor of two or more about the mean price. We show that a calculated power purchase agreement price for the plant improves by up to 10% when operating under California energy market conditions. Sensitivity analysis on the thermal storage cost shows that the optimal design remains unchanged even when doubling costs. [ABSTRACT FROM AUTHOR]

SNCLFR-100 is a small modular natural circulation lead-cooled fast reactor, developed by University of Science and Technology of China, aiming at taking full advantage of the good economics and inherent safety of lead-cooled fast reactors to develop miniaturized, lightweight and multi-purpose special nuclear reactor technology. SNCLFR-100 is still in the conceptual design stage, in order to fully evaluate the safety features of the reactor and provide reference for the optimization design of the next stage, three typical transients are selected based on the analysis of the SNCLFR-100 initiating events by using the code Analysis of Thermal-hydraulics of Leaks and Transients (ATHLET), which are unprotected transient overpower (UTOP), unprotected loss of heat sink (ULOHS) and unprotected partial blockage in the hottest fuel assembly. For UTOP, the unexpected positive reactivity insertion of 0.7$ in 15s led to two large power peaks in the core quickly, and then the core power began to decrease and gradually stabilized under the action of various of negative feedbacks of the reactor, the peak temperatures of fuel and cladding rose rapidly with the increase of core power and eventually stabilized at a higher temperature. For ULOHS, as the reactor were driven by natural circulation, the coolant mass flow rate continued to decline after the transient, both core and cladding temperatures rose quickly and the temperature rise were smaller than that of UTOP transient, the reactor shutdown by itself and the peak temperatures of fuel and cladding were smaller than the safety limit. For unprotected partial blockage in the hottest fuel assembly, with the increase of the blockage rate of the hottest fuel assembly inlet, the coolant flow rate, the peak temperatures of coolant, fuel and cladding increased significantly, when the blockage rate increased to 0.9, the coolant flow rate of the hottest fuel assembly dropped to about 12.6% of the normal value, and the cladding peak temperature would exceed the cladding melting point, measures should be taken to avoid the happening of severe accident.

• VOF simulations of water jet injection into high-temperature melt are performed. • Validation on small-scale tests for water jet – LBE melt pool interaction is performed. • All experimental stages of cavity evolution in the melt pool are reproduced. • Good agreement with multi-fluid simulations are demonstrated. • Heat transfer coefficient agrees with the experimental range for film boiling regime. Numerical study of water jet impact on a pool of high-temperature metallic melt is carried out. Experimental studies in this field encounter difficulties in visualizing the flow and performing measurements in hot and opaque medium, while theoretical studies by multifluid models have severe limitations in describing the flow structure and interphase processes. VOF method allows one to alleviate many limitations, as well as reproduce the flow details that are hardly amenable to direct experimental measurements. The work is focused on the validation of VOF predictions against small-scale experimental data on water jet interaction with molten lead–bismuth eutectic alloy; also, features of the vapor-water cavity formation and evolution in the pool are analyzed. Simulations are performed by a solver developed in OpenFOAM software. Results are presented for three-dimensional simulations of subcooled water jet interaction with LBE melt in a vessel, which has a 10-mm thick slice configuration with semi-circular bottom. Validation is performed against two tests by Sibamoto et al, for a water jet impacting the melt at the velocity of 4.7 and 6.2 m/s. Cross-verification with two multi-fluid codes is performed. Reasonable agreement with respect to cavity geometry and evolution stages is demonstrated, further research directions are discussed. Results obtained are relevant to nuclear safety problems for advanced lead-cooled fast reactors. [ABSTRACT FROM AUTHOR]

Variable renewable energy availability has increased the volatility in energy prices in most markets. Nuclear power plants, with a large ratio of capital to variable costs, have historically operated as base load energy suppliers but the need for more flexible operation is increasing. We simulate the techno-economic performance of a 950 MWt nuclear power plant, based on the Westinghouse lead-cooled fast reactor, coupled with molten salt thermal storage as a method for flexible energy dispatch. We use the System Advisor Model to model the nuclear reactor thermal power input and power cycle operating modes. We combine this robust engineering model with a mixed-integer linear program model for optimized dispatch scheduling. We then simulate the coupled nuclear and thermal storage system under different market scenarios with varying price volatility. We find that the coupled plant outperforms the base plant under markets where energy price peaks fluctuate by a factor of two or more about the mean price. We show that a calculated power purchase agreement price for the plant improves by up to 10% when operating under California energy market conditions. Sensitivity analysis on the thermal storage cost shows that the optimal design remains unchanged even when doubling costs.

BackgroundIn order to accurately predict the heat transfer characteristics in the sub-channel of fuel assembly for the low Prandtl number fluid, it is necessary to select a suitable turbulent-Prandtl-number model.PurposeThis study aims at the applicability of turbulent-Prandtl-number models for liquid lead-bismuth turbulence in triangular fuel rod bundles.MethodsBased on the triangular rod bundle heat transfer correlation, the turbulent models of Shear Stress Transfer (SST) k-ω was adopted to analyze the different rod bundle structures for five different kinds of turbulent-Prandtl-number models. The analysis results were compared with the recommended heat transfer correlation, which had been verified by experiment in the sub-channels for liquid lead-bismuth. The applicability of various turbulent-Prandtl-number models under different rod diameters and pitch diameter ratios was analyzed in details.ResultsThe results show that the global turbulent-Prandtl-number model is not only related to the Reynolds number Re and Peclet number Pe, but also related to the pitch-to-diameter ratio (P/D). In the entire region of (1.3≤P/D≤1.7), simulation results of turbulent-Prandtl-number model proposed by Kays are in good agreement with the values calculated by Mikityuk equation. All sorts of turbulent-Prandtl-number models have the best P/D scope of application.ConclusionsThe related turbulent-Prandtl-number models can be used to predict the heat transfer characteristics of liquid lead-bismuth in the triangular rod bundles under different pitch diameter ratios.

Summary: The lead‐cooled fast reactor (LFR) offers enhanced safety and reliability with the fine properties of liquid lead and lead alloy. To study accurately the thermal characteristics of fast reactors, the multiscale thermal‐hydraulic coupling simulation is an effective way. Multiscale coupling based on the sub‐channel code has evident advantages on the analysis of fuel assemblies. In this study, a multiscale thermal‐hydraulic analysis of a forced‐circulation, medium‐power LFR under steady‐state and transient conditions is performed with the system code ATHLET and sub‐channel code KMC‐SUBtraC which was developed based on the previous version by modifying the pressure drop correlations and adding the assembly‐level calculation. The codes are one‐way‐coupled, with good efficiency and precision. Transient verification of the sub‐channel code is conducted with the CFD code. In the steady‐state analysis of M2LFR‐1000, mass flow and temperature distributions of the assemblies, sub‐channels, and fuel rods in the hottest assembly are analyzed and the safety performance is investigated. In the transient analysis, two typical DECs (unprotected overpower transient and ULOF+ULOHS) are simulated and the multiscale thermal‐hydraulic characteristics are analyzed. With the negative reactivity feedback, the variations of the temperatures of the coolant and fuel rods are within the safe limits, which shows the inherent safety of the reactor. And the results indicate that the loss of primary flow could increase the risk of cladding corrosion. [ABSTRACT FROM AUTHOR]

The Belgian Nuclear Research Centre (SCK•CEN) is at the forefront of heavy liquid-metal (HLM) nuclear technology worldwide with the development of the Multi-purpose hYbrid Research Reactor for High-tech Applications (MYRRHA) accelerator-driven system. MYRRHA is a flexible fast-spectrum pool-type research reactor cooled by lead bismuth eutectic (LBE) and has been identified as the European Technology Pilot Plant for the lead-cooled fast reactor. Given the innovative nature of MYRRHA, the project is currently going through a prelicensing phase. The MYRRHA research and development (R&D) program is driven by this prelicensing process and aims to fill the existing gaps in knowledge with respect to LBE chemistry, material behavior, fuel behavior, instrumentation, and HLM thermal hydraulics. In this critical review we present selected topics from the R&D program on HLM thermal hydraulics that are essential for the design, engineering, and safety analysis of MYRRHA and other HLM-cooled reactors. The topics addressed include turbulent heat transfer in HLM, fuel assembly thermal hydraulics and flow-induced vibrations, control rod hydrodynamics, primary heat exchanger pool and integral system thermal hydraulics, sloshing, and multiphysics modeling. [ABSTRACT FROM AUTHOR]

• The jet interaction between hot and cold fluids is accompanied by strong heat exchange and phase change processes, and such dramatic variations lead to the eventual dispersion of numerous fragmented bubbles in the high-temperature liquid pool. • The test result shown that local vapor explosion is the primary mechanism leading to bubble fragmentation. • The fragmented bubble size distribution was modeled, which is promising to provide guidance for the bubble behavior characteristics after steam generator tube rupture accidents in lead-cooled fast reactors. Bubble fragmentation and size distribution in hot and cold fluid jet interaction are fundamental for industries such as energy and metallurgy, especially for safety analysis after steam generator tube rupture (SGTR) accidents in lead-cooled fast reactors. The opaqueness to light of the molten lead pool makes it difficult to capture the gas-liquid interface behavior and thus lacks relevant data. In this paper, silicone oil was used instead of lead to simulate the injection process of the SGTR accident. Then the jet visualization experiments were carried out to investigate the bubble fragmentation characteristics. The visualization results show that the jet between the hot and cold fluids process is accompanied by strong thermal interactions, eventually leading to many bubbles dispersed in the high-temperature liquid pool. Secondly, four mechanisms of action affecting bubble size were elaborated in this paper, among which local vapor explosion is the primary mechanism of bubble fragmentation. Finally, the data revealed that bubble size presented a lognormal distribution, and the fragmented bubble size distribution was modeled. This study is promising to provide guidance for the bubble behavior characteristics after SGTR accidents in lead-cooled fast reactors. [ABSTRACT FROM AUTHOR]

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

The lead-cooled fast reactor (LFR) coupled with SCO 2 Brayton cycle is an advanced nuclear power generation system which has high efficiency and high compactness. The purpose of this study is to investigate the impact of coupling effect between LFR and SCO 2 cycle on the safety analysis of unprotected transient overpower (UTOP) accident, and to propose mitigation measures to UTOP accident with the aid of SCO 2 cycle. Firstly, the dynamic simulation model of a 100 MW small modular natural circulation LFR coupled with SCO 2 Brayton cycle was developed using APROS software, and was verified through the steady-state and transient simulations. Secondly, the UTOP accident of the LFR coupled with SCO 2 Brayton cycle was analyzed, including a comparative case which simplified SCO 2 cycle as a fixed boundary in the analysis. Finally, two measures to mitigate the UTOP accident by charging more SCO 2 to the circulation loop from the storage tank and by maintaining constant SCO 2 temperature at the inlet of in-reactor heat exchangers (IRHXs) were proposed. The results indicate that the conventional method of simplifying power generation system as the fixed boundary underestimated the severity of accident. Specifically, the maximum fuel cladding temperature was about 40 °C higher when considering the coupling effect of reactor and SCO 2 cycle than that of assuming SCO 2 cycle as the fixed boundary, which greatly exceeded the safety limit. Moreover, within the scope of present UTOP accident conditions, the measures proposed in this study could mitigate the severity of accident. The maximum cladding temperature was reduced by 15–23 °C in the condition of maintaining constant SCO 2 temperature at IRHX inlet, and was reduced by 5–7 °C in the condition of charging more SCO 2 to the circulation loop, compared with the condition not applying any measures. [ABSTRACT FROM AUTHOR]

• The steam transport behavior was simulated in a postulated SGTR accident of LBE-cooled reactor. • Steam bubble transport and the influence factors including the failure location and pressure of secondary loop were analyzed. • For the tube failure located at the outlet of SG, the maximum volume fraction of steam transported to reactor core was higher than the middle and inlet conditions. • The maximum steam volume fraction in reactor core increased with the increasing of the pressure of secondary loop. The multiphase flow phenomenon of the interaction between lead-bismuth eutectic and water is a safety concern for the design of lead-based fast reactor. In order to investigate the steam transport characteristics under a postulated Steam Generator Tube Rupture (SGTR) accident in a lead-based reactor, numerical simulation of lead-bismuth eutectic-water/steam multiphase flow in a reactor pool was performed and the effects of failure location and pressure of secondary loop on steam volume fraction in reactor core was evaluated. Under the present theoretical analysis, the results showed that the maximum steam volume fraction in the core occurred for the condition of cooling tube failure located at the outlet of SG which is closest to the pump duct and the accident consequence might be the most serious. The maximum steam volume fraction in reactor core increased with the increasing of the pressure of secondary loop. The analysis results will provide reference for the safety design and engineering guidance of lead-cooled fast reactor. [ABSTRACT FROM AUTHOR]

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