Search

• SACOS-PLATE: Enhanced code for plate fuel analysis. • Validated against RA-6, Bettis experiments. • Unique heat conduction, flow models. Over the past few decades, plate-type fuel elements have been widely used in advanced research reactors, small modular pressurized water reactors, production reactors, etc., with the advantages of compact structure arrangement, good heat transfer characteristics and high fuel consumption. Currently, a thermal–hydraulic subchannel analysis code, namely SACOS-PLATE code, was newly designed to satisfy the calculation of different types of plate fuel elements under various steady-state operating conditions. Based on the original SACOS code, this updated version modifies the flow distribution model, the plate heat conduction model, the heat distribution model as well as the flow heat transfer model, and then realizes the simulation of the special thermal–hydraulic phenomena of the fuel plate and the rectangular flow channel. In this present study, the RA-6 and Bettis experimental data was used to validate the code, and the simulation calculation was performed on the JRR-3 research reactor, which effectively proved the good computing capability of SACOS-PLATE code. [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]

• Combining the compactness and mobility of heat pipe reactors, a mobile nuclear-electric hybrid energy storage system based on the heat pipe-cooled reactor has been proposed for the first time. • The safety of the core under different reactor accident conditions is analysed based on a neutronics-thermal coupling analysis approach. • The design values of the energy storage system were systematically analysed for the target parameters. To achieve the goals of carbon emission peak and carbon neutrality, it is necessary to expand support for non-fossil energy sources. Heat pipe reactor (HPR) is a new reactor design concept that uses the efficient, passive thermal conductivity of heat pipes to cool nuclear fuel, which makes solid state HPR very suitable as a power source for mobile transport vehicles. However, it is also faced with the problem of frequent start-stop and variable output. In this paper, a conceptual design of nuclear power and energy storage coupled power system was proposed, using uranium nitride as fuel placed in a solid metal matrix, with the sodium heat pipe as the heat conduction device to remove heat from the core. The condenser of the heat pipe is connected to a dual-mode power generation system, comprising a thermoelectric generator and a Stirling generator. Additionally, a hybrid energy storage system, consisting of power battery packs and supercapacitors, is integrated to facilitate power output adjustment and energy storage. This comprehensive configuration enables the realization of a peak power output of 350 kW. Finally, the neutronics and thermal characteristics of the design were analyzed. Calculations showed that the radial power peak factor was 1.2396 and 1.4924 for the axial direction, it also showed the maximum core temperature to be 1269.6 K under steady state conditions with neutronics and thermal coupling, which is 16.54 % higher than before coupling. [ABSTRACT FROM AUTHOR]

The isothermal steam oxidation behavior of Cr-coated Zr-1Nb alloy at 1350 ℃ and 1400 ℃ was investigated. The temperatures are above the Cr-Zr eutectic temperature. The reliability of the experimental setup and methodology was confirmed by conducting isothermal steam oxidation experiments on bare Zr-1Nb specimens at 1100∼1400 ℃, and the corresponding oxidation kinetics correlation was obtained. After the eutectic-oxidation reaction, the surface of the coated specimens exhibited a "crocodile skin" morphology composed of "ridges" and "hollows", which was attributed to the non-uniform eutectic reaction. Due to the much higher O affinity of Zr compared to Cr, and the higher solubility of Cr in β-Zr compared to α-Zr(O) and ZrO 2 , the ZrO 2 and α-Zr(O) layers initially grew on the Cr-Zr eutectic during oxidation. Cr continuously diffused inward and eventually precipitated as Cr-rich phases within the metallic matrix. Three types of Cr-rich phases were observed, with varying oxygen content and reflecting the oxidation process of the Cr-rich phases. The Cr-rich phases were ultimately oxidized into a mixture of Cr 2 O 3 and ZrO 2. The parabolic weight gain rates for the oxidation of Cr-Zr eutectic are compared with the oxidation kinetics correlation derived from the bare Zr-1Nb specimens, and the model applicability of the oxidation kinetics correlations of Zr alloys in the simulations of Cr-Zr eutectic-oxidation reaction is discussed. [ABSTRACT FROM AUTHOR]

The lead-bismuth cooled fast reactor core adopts box-type design, and the fuel rod bundle is fixed by wrapping wires. The complex geometric structure of the fuel assemblies significantly affects the flow and heat transfer status of the liquid metal coolant. Building upon the open-source CFD platform, OpenFOAM, this work established conservation equations for liquid metal coolant, fuel rod thermal characteristic models, inter-wrapper coolant analysis models, and coupled heat transfer models, and developed the subchannel level thermal-hydraulic analysis code CorTAF-LBE for whole-core analysis of LBR core using the finite volume method. The wire-wrapped rod bundles internal flow and heat transfer experiments of lead-bismuth, carried out based on the THEADES and KYLIN-II circulation loops, were selected to perform verification and validation. The results match well with the experimental data and CFD fine-modeling prediction. The steady-state numerical simulation of three wire-wrapped rod bundle assemblies with inter-wrapper flow was carried out under normal operation condition. The temperature and pressure distributions of the coolant in each subchannel and inter-wrapper flow region, the temperature of cladding and other key thermal-hydraulic parameters were obtained. The role of the inter-wrapper coolant in heat removal from the fuel assemblies was revealed. The work in this paper has referential value for the optimization of lead-bismuth cooled fast reactor design and safety assessment during operation. [ABSTRACT FROM AUTHOR]

The liquid phase entrainment through ver tical-up branch had been analyzed through scaled and full size of experiment since ADS-4 (the fourth-stage Automatic Depressurization System) valve adopted in advanced PWR (Pressurized Water Reactor). Most of the existed studies focus on the single inlet of gas whereas steam may flow from two sides of the T-junction during the blowdown of primary system. Hence the experimental study on liquid entrainment with double end inlets are carried out. The data on the onset of entrainment (OLE) turns out that double end inlets show small difference with single inlet, whereas the entrainment rate research have great difference with single inlet and there is more than 50% error between the data and existing theoretical models. According to the visible flow phenomenon and flow instability, a new entrainment rate model was proposed which can well predict the double-end inlets and single-inlet entrainment phenomenon with the error within ±20%. • Entrainment experiments with double inlets in different flow rate were conducted. • The applicable test methods are discussed and new phenomena and data are obtained. • A new entrainment rate model with smaller prediction error is proposed. [ABSTRACT FROM AUTHOR]

Sodium-cooled fast reactor (SFR) is one of the most promising Generation IV reactor types. To improve operational safety and economy, SFR requires accurate prediction of the steam power conversion system. Therefore, a one-dimensional thermal-hydraulic analysis code of the steam power conversion system of SFR is developed in this study where each equipment is packaged as a separate module with modular modeling method. JFNK is adopted to solve ODE in this code, which expand the universality of the code. The model is validated by comparing with design results from the Daya Bay nuclear power plant and the BN-600 steam generator. The maximum relative errors for pressure and temperature are 5.44% and 2.40%, respectively. After validation of the code, the steady-state analysis of the SFR steam power conversion circuit is carried out where the flow and heat transfer process of key equipment can be accurately simulated considering a complete steam power conversion cycle. This work provides reference value for the thermal-hydraulic characteristics of steam power conversion system in SFR. • A thermal-hydraulic analysis model of the steam power conversion system of SFR is developed by using C++ language with modular modeling method. • The model is validated with design results of the Daya Bay Nuclear Power Plant and the BN-600 steam generator. • The steady-state analysis of the SFR steam power conversion circuit is carried out. [ABSTRACT FROM AUTHOR]

Abstract: A Water-cooled Pressure Tube Energy production blanket (WPTE) for fusion driven subcritical reactor has been designed to achieve 3000MW thermal power with self-sustaining tritium cycle. Pressurized water has great advantages in energy production; however the high pressure may cause some severe structural design issues. This paper proposes a new concept of water-cooled blanket. To solve the problem of the high pressure of the coolant, the pressure tube was adopted in the design and in the meantime, the thickness of the first wall can be significantly reduced as result of adopting pressure tube. The numerically simulating and calculating of temperature, stress distribution and flow analyses were carried out and the feasibility of using water as coolant was discussed. The results demonstrated the engineering feasibility of the water-cooled fusion–fission hybrid reactor blanket module. [Copyright &y& Elsevier]

Abstract: An integral pressurized water reactor (IPWRs) is a class of reactors in which the main equipments of the primary loop system, core, steam generators (SGs), pressurizer (PZR), and main coolant pumps (MCPs) are housed in an integral reactor pressure vessel (RPV). In this study, a point reactor neutron kinetics model depended on the point kinetics with six groups of delayed neutron, a drift-flow once-through steam generator model, a nitrogen pressurizer model and other auxiliary models were established. A visual transient analysis program named VTSAS 1.0 has been developed to analyze the thermal–hydraulic characteristics of the IPWRs. Interactive visual input, real-time plotting and comparing functions and interactive visual output were implemented using mixed programming of Visual C++ language and FORTRAN90 language, which greatly facilitate applications in the engineering. The VTSAS 1.0 was applied to analyze the transient thermal–hydraulic characteristics of reactivity insertion accident (RIA) and sudden increase of feed-water temperature accident of an IPWR. The results obtained were found to be in reasonable agreement with those by RELAP5/MOD3. The research is significant to the improvement of design and safety operation of the IPWRs. [Copyright &y& Elsevier]

• The basic theory of improved quasi-static methodhas been studied in detail. • Provide the derivation of the first-order perturbation theory of neutron diffusion equation and kinetics parameters. • Implementing this method into the state-of-the-art FVM framework – OpenFOAM, and elaborating on the numerical details. To analyze the transient problems for neutronics and make the preparation for high-fidelity multi-physics simulation. In this study, a neutron space–time kinetics solver called "KaiyangIQS" with the improved quasi-static method based on OpenFOAM is developed. The basic principles of improved quasi-static method, numerical method, and solver strategy are derived and discussed in detail. Separate validations for the modules of solving amplitude function and adjoint scalar flux diffusion equation are carried out. The results show that OpenFOAM can effectively handle the stiff matrix of amplitude equations with high accuracy and robustness. The adjoint scalar flux diffusion equation solver is also validated by the IAEA 3D benchmark. For the improved quasi-static method solver, unidimensional rod ejection problem, TWIGL benchmark problem, and three-dimensional rod ejection problem are carried out for calculation and validation. Normalized power prediction, reactivity prediction, effect of adopting different discretization schemes of transient terms, rod cusping effect, and lag effect of delayed neutron precursors are analyzed in detail. The numerical results of the solver with improved quasi-static method that developed in this study are in good agreement with references, and this solver can be considered a promising tool to simulate the transient neutron kinetics problems. [ABSTRACT FROM AUTHOR]

Pressurizer is widely used in the industrial systems, which can maintain the pressure of the system in a certain range. Especially, as the system pressure rises, the spray droplets interacts with the saturated steam in pressurizer to reduce system pressure. In this paper, a heat and mass transfer model of droplet and steam in pressurizer is established through the lattice Boltzmann method (LBM), and the characteristics of moving droplet in steam space with different Reynolds numbers are analyzed. In this paper, these results show that the flow fields on both sides of the droplet have obvious symmetry at low Reynolds number. As the Reynolds number increasing, the wake region of flow field for moving droplets gradually loses symmetry. When the Reynolds number reaches 103 orders of magnitude, the droplet motion begins to appear horizontal deviation. The growth of droplet condensation radius is inhibited with the increase of Reynolds number. When Reynolds number reaches 2600 and Fo = 0.2, the growth of droplet radius stops and then decreases. Due to the influence of inertia force generated by internal circulation flow and droplet movement, as the Reynolds number increasing, the droplet heat transfer rate decreases at first, but begins increasing when Reynolds number raises to about 2000. The calculation results have reference value for the development of spray droplet model of pressurizer. • A heat and mass transfer model of droplet and steam in pressurizer is established. • The characteristics of moving droplet in steam under different Re are analyzed. • With the increase of Re , the growth of droplet condensation is inhibited. • Heat transfer rate decreases first and then increases with the increase of Re. [ABSTRACT FROM AUTHOR]

The conceptual design of China Fusion Engineering Test Reactor (CFETR) has been completed and is currently in the engineering design phase. In this paper, focusing on three categories of candidate blanket designs and divertor designs, the selection schemes for materials of plasma-facing components, solid structure, tritium breeding materials, neutron multiplier, coolant, and divertor were extracted. Thermo-physical property data was compiled to form a material thermo-physical property database, named Thermo-physical Property Database (ThePD), which is the first advanced thermo-physical database for materials of CFETR blankets. Besides, the callable code was generated, which can be directly applied in the system code for the blanket. Based on internationally available database from National Institute of Standards and Technology (NIST) and similar software AP1700, the correctness of the data and the transitional analysis of regions designed by International Association for the Properties of Water and Steam (IAPWS) were verified. Using the thermal-hydraulic system analysis code, integrated validation was performed for the validated thermo-physical property data and code, demonstrating the applicability and accuracy of the data. ThePD is conducive to addressing the issue of incomplete thermo-physical property data faced by blanket design and safety analysis software for CFETR, and it can fill this gap. [ABSTRACT FROM AUTHOR]

Based on the analysis of energy structure and the development trends of advanced nuclear energy technology, combining the complementary advantages of nuclear energy and other energy sources in China, the Fluoride-Salt-cooled high-Temperature Advanced Reactor (FuSTAR) system has been proposed. FuSTAR system is designed to address production capacity issues in areas with energy shortages or resource concentration and can be combined with high-temperature process heat to achieve high-value-added processes such as hydrogen production. In this paper, the Tri-structural isotropic fuel combined with helical cruciform elements has been studied in detail on the neutron physics and thermal-hydraulic properties, and the flow and heat transfer experiment is carried out for the first time. The multilayer nonlinear programming method has been applied to the design of thermal transport systems. Then the optimal parameters and configuration of the thermoelectric conversion system are obtained by the superstructure optimization method. The inherent safety of FuSTAR is demonstrated through deterministic safety analysis. In addition, the iodine-sulfur cycle technology and the characteristics of tritium production are also analyzed. This study highlights the significant potential of integrating nuclear energy with other energy sources in an optimized energy-production system. • Multi-purpose reactor energy system for power - hydrogen - tritium production. • Over 50 % maximum thermal efficiency. • Simultaneous analysis of design, optimization, and safety. [ABSTRACT FROM AUTHOR]

• A conceptual design of FuSTAR is developed, and validated by neutronics and thermal-hydraulics analysis. • Detailed core design and neutronics analysis with near-critical configuration and burnup calculation are carried out. • Detailed thermal-hydraulics analysis with turbulent flow and heat transfer is carried out. The fluoride-salt-cooled high-temperature advanced reactor (FuSTAR) is an advanced small modular reactor, which aims to the electric supplement and clean heat production to western China. In order to have good economy and safety, FuSTAR needs to have the characteristics of small size, light weight, deep burnup, and inherent safety. A conceptual design of FuSTAR is proposed in this study with neutronics and thermal-hydraulics assessments. In this study, its concept is established with two-region fuel to flatten core power, and the neutron spectrum of the core is in the thermal spectrum to improve the utilization of thermal neutrons and reduce fuel inventory. Neutronics and thermal-hydraulics analyses are also carried out, and established the core layout and material ratio under the near-critical condition. Burnup depth of the fuel, power distribution, neutron flux distribution, flow and heat transfer in rod-bundles channel, and 1/6 core porous media analysis are also investigated. The parameters of the designed core are evaluated quantitatively from various aspects. The results show that the core design in this study has sufficient shutdown depth, the equilibrium cycle is 3.24 years, and the discharge burnup depth is 106 GWd/MtU. The helical cruciform fuel (HCF) adopted by this reactor has a better flow and heat transfer capacity than cylindrical-type fuel at the inner channel. The analysis of 1/6 core porous media shows that the peak fuel temperature under normal operating condition is 1009.3 K, which is far below the limit of 1573 K. Relevant design and analysis results can provide the reference for the subsequent small modular FHR design and construction. [ABSTRACT FROM AUTHOR]

• Tie tube elements greatly improve temperature safety. • The 2:1 support ratio is still the best choice. • Deposition energy cannot be ignored in the SNTP system. • The issue of sudden temperature rise on the outer tie tube is worth noting. As a main structure, tie-tube elements are used both in the traditional and modern LEU nuclear thermal propulsion reactors. This complex structure and its corresponding complex convective heat transfer process have been studied, including CFD calculations. But these studies did not effectively evaluate the applicability of the models and typical problems. In this study, a general small nuclear thermal propulsion (SNTP) reactor model is established by using three-dimensional Monte Carlo code and commercial CFD program Fluent. The applicabilities of material properties, turbulence models and grids are clarified by analyzing various influencing factors. The calculation results show that due to the small enthalpy difference and heat transfer, the convection effect of inner flow based on hydrogen can be ignored. There is no factor that can significantly reduce the maximum fuel temperature, but the porosity of the insulator can significantly improve the heat transfer ratio. In terms of thermal safety and overall performance, 2:1 support ratio is still the best choice. Based on the simulation results, the phenomenon of steep temperature rise of outer tie-tubes is found, and the SST k-omega model with Y + <3 is recommended for simulation of similar problems. [ABSTRACT FROM AUTHOR]

The oxidation and degradation mechanisms of Cr coating on Zr alloy cladding under high temperature steam atmosphere are summarized. Six basic phenomena are considered for model development, including: oxidation of Cr, growth of ZrCr 2 , formation of Zr(Cr) solid solution, outward diffusion of Zr, reduction of Cr 2 O 3 , and volatilization of Cr 2 O 3. The mathematical analysis model is established and validated, demonstrating its good capacity to predict the increase and subsequent decrease in the thickness of Cr 2 O 3 scale, as well as the thickness changes in other individual layers. The developed model is applied in the analysis of the structure evolution of Cr coating under steady state, transient heating conditions and postulated accident scenarios. [ABSTRACT FROM AUTHOR]

• Sliding mesh method was used to simulate moving conditions. • Comparing the numerical results with the experimental results. • Two calculation methods of simulating moving conditions were compared. • It is concluded that the effect of rolling motion is the most sever. • Secondary flows and flow instabilities under moving conditions were analyzed. The investigation of flow and heat transfer of lead-bismuth eutectic (LBE) under moving conditions is vital to the design optimization of LBE loop system in offshore engineering. In this paper, we have developed a method for simulating moving conditions using sliding mesh. Moreover, we have conducted experiments to validate the numerical model, and the comparison results demonstrate a good agreement between them. The calculation results demonstrate that the period of the flow rate fluctuation is consistent with the motion period, and the amplitude of the fluctuation will increase with the increase of the maximum range of motion and the decrease of the motion period. In addition, the influence of rolling on system parameters is more severe than that of heaving. These results obtained in this study have important engineering application value for the design of LBE loop system under moving conditions. [ABSTRACT FROM AUTHOR]

• Subchannel code for the thermal hydraulic analysis of liquid metal cooled reactor. • Simulation of transition from forced to natural circulation by subchannel code. • Effect of non-uniform heating on the thermal field. • Validation of YAOGUANG code under steady and transient conditions for LBE coolant. Based on the experimental campaign performed on the NACIE-UP (NAtural Circulation Experiment-Upgraded) facility, the benchmark activity was proposed by ENEA to qualify and validate the numerical tools for the Heavy Liquid Metal (HLM) system. Xi'an Jiaotong University participated in the benchmark exercise of subchannel validation and the self-developed subchannel code YAOGUANG was applied to the thermal hydraulic analysis. The performed two tests with different power distributions on the Fuel Pin bundle Simulator (FPS) were simulated, characterized by transition from forced to natural circulation. The applicability of this code to the stable state and mass flow rate transient was evaluated by comparing the numerical outcomes with the experimental results. Then the effect of the non-uniform heating on the thermal field was analyzed. The results indicate that this subchannel code has good accuracy in simulating both steady state and transient conditions with uniform power distribution. For the non-uniform case, the subchannel calculations can capture the general trend of temperature distributions, with a larger simulation deviation observed. This discrepancy is mainly due to the limitations of existing models adopted because they didn't consider the impacts of non-uniform conditions on the flow and heat transfer in rod bundles. Finally, the influence of non-uniform heating is mainly manifested as the increase of axial and radial temperature gradients and it helps to flatten the radial thermal field to some degree. [ABSTRACT FROM AUTHOR]

Thermal stratification is an important phenomenon in the process of natural circulation decay heat removal in sodium-cooled fast reactors (SFR). It affects the establishment of the natural circulation and the flow rate of the natural circulation in a station black out (SBO) accident, and therefore affects the safety of the reactor. In safety evaluation, programs are often required to perform analysis for various conditions and long-term transients. However, it is difficult to meet this requirement by adopting a multi-dimensional method or one-dimensional and multi-dimensional coupling method. Therefore, this paper proposes applying a thermal stratification model to the system analysis program THACS to analyze the thermal stratification phenomenon. First, the thermal stratification model is validated with experimental data. The validated model is applied to analyze a station black out accident in the SFR. The lumped parameter model and thermal stratification model are used respectively for the hot pool. By comparing the calculation results of the two models, the effect of thermal stratification of the hot pool on key parameters such as the natural circulation flow rate of the primary circuit, natural circulation flow rate in the inter-wrapper region, average core outlet temperature, maximum core fluid temperature, and the maximum temperature of cladding was analyzed, and the effect of thermal stratification on natural circulation decay heat removal of the SFR was evaluated. This work can provide a reference for the safety analysis of the reactor and contribute to developing a more conservative safety analysis and evaluation strategy. [ABSTRACT FROM AUTHOR]

Abstract: The steam generator (SG) secondary water-cooled and air-cooled passive residual heat removal systems (PRHRSs) are proposed and designed for 300MW Pressurized Water Reactor (PWR). The RELAP5/MOD3.4 code is utilized to study the behavior of the PRHRSs and transient characteristics of primary loop system during Feed-water Line Break (FLB) accident. The characteristic comparison between water-cooled and air-cooled PRHRSs is also conducted in this study. The results show that both water-cooled and air-cooled PRHRSs can establish stable natural circulations in PRHRS loops, which realize the effective core decay heat removal from primary loop. Results illustrate that both water-cooled PRHRS and air-cooled PRHRS designed in the study have great significance for improving the inherent safety of 300MW nuclear power plant. However, the water-cooled PRHRS heat transfer ability is stronger in the initial time, while it becomes weaker as the system tank water evaporates in the later stage. In contrast, the air-cooled PRHRS, ensuring the main reactor operation parameters to decrease to a more safety value, is more effective than water-cooled PRHRS for long term cooling. [Copyright &y& Elsevier]

Abstract: TACOS(Transient Analysis Code Of SCWR) code has been developed to analyze the safety characteristics of SCWR-M for present work. In the code TACOS, the single channel model, point kinetic model, heat transfer model and control system model were used. Four different heat transfer correlations were applied to the steady state calculation of SCWR with mixed spectrum core (SCWR-M). After the comparison of four correlations, Bishop Correlation was selected for transient analysis. Using TACOS code, preliminary analysis of loss of flow accident (LOFA) for SCWR-M has been performed. Partial loss of flow and total loss of flow accident were discussed respectively. For the total loss of flow, the results indicate that the maximum cladding surface temperature (MCST) in the thermal zone was much higher than that in the fast zone. A “special flow mode” similar to natural circulation between coolant channels and moderator channels in the thermal zone has been found. The co-current flow mode design in the thermal zone would cause disadvantage to safety performance because it has a negative effect on the flow rate of coolant channels after the startup of Auxiliary Feedwater System (AFS). The sensitivity analyses of pump coastdown time and AFS flow rate for total loss of flow accident were also carried out. [Copyright &y& Elsevier]

Abstract: The Pebble Bed Water-cooled Reactor (PBWR) is a water-moderated water-cooled pebble bed reactor in which millions of tristructural-isotropic (TRISO) coated micro-fuel elements (MFE) pile in each assembly. Light water is used as coolant that flows from bottom to top in the assembly while the moderator water flows in the reverse direction out of the assembly. Steady-state thermal–hydraullic analysis code for the PBWR will provide a set of thermal hydraulic parameters of the primary loop so that heat transported out of the core can match with the heat generated by the core for a safe operation of the reactor. The key parameters of the core including the void fraction, pressure drop, heat transfer coefficients, the temperature distribution and the Departure from Nucleate Boiling Ratio (DNBR) is calculated for the core in normal operation. The code can calculate for liquid region, water-steam two phase region and superheated steam region. The results show that the maximum fuel temperature is much lower than the design limitation and the flow distribution can meet the cooling requirement in the reactor core. As a new type of nuclear reactor, the main design features with a sufficient safety margin were also put forward in this paper. [Copyright &y& Elsevier]

Abstract: In this study, we have developed a thermo-hydraulic and safety analysis code named TSAC1.0 with Visual Fortran 6.5 to analyze the thermal-hydraulic characteristics of the China advanced research reactor (CARR) under reactivity insertion accident (RIA) which was induced by unexpected control rod withdrawal in full power condition. The neutron kinetic model depended on the point kinetics with six groups of delayed neutrons including reactivity feedback effects and it was adopted for the solution of reactor power. Furthermore, a new simple and convenient model was adopted for the solution of the transient behaviors of main pump instead of the complicated four-quadrant model. Visual input, real-time processing and dynamic visualization output were achieved using Microsoft Visual Studio.NET 2003 to make the application of TSAC1.0 much more convenient in the engineering. The simulated results of TSAC1.0 were found to be in reasonable agreement with those of RELAP5/MOD3 and showed that the parameters, including the peak coolant temperature, the peak heat structure temperature, and MDNBR, were in the acceptable range of design safety limit under RIA. [ABSTRACT FROM AUTHOR]

Abstract: In this study, the 3D flow and heat transfer characteristics in rod bundle channels of the super critical water-cooled reactor were numerically investigated using CFX codes. Different turbulent models were evaluated and the flow and heat transfer characteristics in different typical channels were obtained. The effect of pitch-to-diameter ratio (P/D) on the distributions of surface temperature and heat transfer coefficient (HTC) was analysed. For typical quadrilateral channel, it was found that HTC increases with P/D first and then decreases significantly when P/D is <1.4. There exists a “flat region” at the maximum value when P/D is 1.4. If P/D is larger than 1.4, heat transfer deterioration (HTD) occurs as main stream enthalpy is quite small. Furthermore, the HTD under low mass flow rate and the non-uniformity of circumferential temperature were also discussed. [ABSTRACT FROM AUTHOR]

Summary: Heat pipe cooled reactor is designed to meet the future energy demands with many unique advantages such as modularization, solid and static state, high reliability and passive safety. In the innovative nuclear reactor design, heat pipes are used to cool the reactor core and transfer the fission heat passively, thermoelectric generators (TEGs) are utilized to achieve static energy conversion. In this article, a finite element model is established to investigate the detailed TEG thermoelectric performance coupled with heat pipes. A temperature field and flow field coupling model is used to simulate the heat transfer performance of the potassium heat pipe including a heat pipe wall, wick and vapor region. A heat pipe TEG experimental setup is established and the steady‐state experiment is conducted. The experimental setup is selected as the numerical simulation object and the simulated results are compared with the experimental data to verify the numerical model. The maximum simulation error is within 5%. Based on the verification and validation, the steady‐state thermoelectric performances, the effect of contact thermal resistance and external load on the skutterudite TEG thermoelectric performances are investigated. The detailed heat flux distributions, temperature distributions and potential distributions are researched in the steady‐state simulation. The temperature distributions of the TEG cross section are centrally symmetrical and the horizontal temperature difference is 26°C. The temperature difference introduced by thermal resistance at the TEG hot side increases from 33°C to 94°C with the increasing heat flux through TEG module. The influence of thermal resistance on TEG hot side is more significant than that on TEG cold side. The effect of external load on TEG output characteristics is investigated under different heat flux conditions. With the increase of heat flux from 60 to 166 kW/m2, the temperature difference of TEG module and p‐n unit increase by 432°C and 348°C, the thermoelectric conversion efficiency is improved from 3.6% to 8.5%, and output power increases by 20 W, respectively. The research results are of guiding significance for exploring the application potential of TEGs in the static heat pipe cooled reactor. This work provides research foundations for the multiphysics coupling analysis of the micro heat pipe cooled reactor. [ABSTRACT FROM AUTHOR]

Summary: Numerical simulation of the Space Reactor Power System is of great importance to reactors development and application in aerospace. In this paper, thermal‐hydraulic and thermoelectric characteristics analysis based on 1/6th full thermionic reactor core of the TOPAZ‐II were performed using 3‐D computational fluid dynamics software STAR‐CCM+. A Benchmark study was conducted by comparing the numerical results with experimental data and design values with a maximum error of 5.4%‐11.8%. For the thermal‐hydraulic analysis: (a) the highest temperature point occurred at the center hole of the center thermionic fuel element (TFE) with the temperature of 2303 K; (b) the maximum temperature difference located in the fuel region and interelectrode gap with the value about 200 K and 1000 K, respectively; (c) axial temperature distribution of coolant is similar in each flow channel, and the maximum temperature difference appears at the reactor core outlet with the value of 10 K. For the thermoelectric analysis, the potential on center or the edge TFE is much higher or lower than the others, the calculated electrical power of the full thermionic reactor core is 4.95 kW, the distribution trend of current and potential is in a good agreement with the axial emitter temperature. [ABSTRACT FROM AUTHOR]

Summary: Closed type assembly leads to complicated core thermohydraulics which are significant for design and safety analysis in sodium‐cooled fast reactor (SFR). Most of the major phenomena can be dealt with by general parallel channel network and two‐dimensional model for inter‐wrapper flow (IWF) in system code transient thermal‐hydraulic code for analysis of SFR (THACS). A main deficiency left is the overestimation of inter‐assembly heat transfer effect due to lump temperature message that one‐dimensional assembly model can only provide. A methodology was thus developed to compute wall‐subchannel temperature for simulating energy transportation through duct wall. Intra‐assembly thermohydraulics were classified as two mechanisms and non‐dimensional parameter was introduced for characterizing each of them. Ratio between linear heat flux density through hexagonal wrapper surface and pin surface qw/p was employed to account for radial heat transfer. Flow redistribution effect arouse in mixed and natural convection was concretized by a importance symbol of buoyancy force, Richardson number Ri. Two criteria were also proposed for dividing flow state into three parts and specific model for each section was defined. Two validation cases against 3‐D computational fluid dynamics (CFD) simulation of China experimental fast reactor whole core and test data from plant dynamics test loop with direct heat exchanger experiment were performed with THACS. Noticeable improvement after incorporation of this model demonstrated its value in system analysis. Although the procedure for parameter computations is geometry related, this method can be reproduced to any type of assembly directly. [ABSTRACT FROM AUTHOR]

Summary: In this paper, development and preliminary validation of an in‐house code LETHAC (lead‐based fast reactor thermal‐hydraulic analysis code) is introduced. LETHAC is developed for predicting the thermal‐hydraulic transient performance of a lead‐bismuth reactor. The mathematical models and module relations of LETHAC are presented in detail. The most‐used LBE (lead‐bismuth eutectic) flow and heat transfer correlations are compared against data from KYLIN‐II 61‐rod wire‐wrapped bundle experiments. Results show little difference between the chosen heat transfer correlations as the relative error of these correlations is all within 2%. Novendstern correlation is recommended for wire‐wrapped bundle frictional resistance coefficient calculation as it fits the experimental data best. Besides, the transient calculation capability of LETHAC is also preliminarily evidenced by TALL transient tests. Good agreements are shown between the calculated temperatures and the experiment ones. Relative errors between calculation results and experimental data are almost within 10% during the whole transient process, which indicates LETHAC can be used for predicting the transient behavior of lead/lead‐bismuth cooled systems. [ABSTRACT FROM AUTHOR]

Summary: To evaluate residual heat removal capacity of next generation China sodium‐cooled fast reactor (SFR) and provide data for code validation, an integral natural circulation experiment was performed on a scaled water platform which had same structure with reactor primary system. A modular code THACS focusing on safety analysis of SFR was validated by this test. Some necessary development and verification for physical property, frictional loss, and heat transfer of water were carried out first. Whole simulation of an unprotected loss‐of‐flow (ULOF) case was divided into two phases and computation lasted until 5000 s of test time during second stage. Agreement in light of system temperature distribution in first stage provided a suitable initial condition for subsequent transient. Satisfied flow rate curve was reproduced with decaying pressure head and valve characteristic as input when pump was stopping. All natural circulation paths were predicted, including original flow in primary loop through main pipe and reversed flow in reactor vessel cooling system (RVCS). Transition to operative mode of direct heat exchanger (DHX) in direct reactor auxiliary cooling system (DRACS) and its cooling power were also forecasted successfully. This work means a significant advance for THACS in the field of engineering application of China SFR. [ABSTRACT FROM AUTHOR]

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: Steam generator is the heat transfer component of sodium‐cooled fast reactor (SFR). It is possible to happen flow instability in the tubes of sodium heated once‐through steam generator (OTSG). In order to investigate the flow instability of parallel channels in sodium heated OTSG, a numerical analysis code called COSFIAS has been developed in Fortran language based on time‐domain method. The experimental data of parallel channels flow instability with uniform heating power condition and sodium heated condition have been applied to validate the code. The relative errors between calculation results and experimental data of uniform heating power condition are within 10%. The relative errors between calculation results and experimental data of sodium heated conditions are within 7%. The comparison results indicate that the COSFIAS program is reasonable and accurate for practical engineering application. In addition, the effects of important parameters on flow instability of parallel channels have been analyzed. From the calculation results some conclusions could be drawn: the decrease of water pressure would cause the increase of heat transfer power and destabilize the parallel channels system; in high subcooling region, the increase of inlet subcooling would stabilize the system; in low subcooling region, with the decrease of inlet subcooling, the parallel channels system would turn to be unstable and then return to be stable again; the increase of water mass flow rate could stabilize the parallel channels system.; the sodium heated condition is more stable than uniform heat flux condition in the same heat transfer power; the increase of sodium inlet temperature could destabilize the parallel channels system; the increase of sodium mass flow rate could destabilize the parallel channels system. In addition, the code could be applied to support the design work of sodium heated OTSG. [ABSTRACT FROM AUTHOR]

Summary: Helium‐Xenon (He‐Xe) gas mixture featured with low Prandtl number is increasingly used in various energy devices as heat‐carrying agent. To figure out the heat transfer characteristics of He‐Xe gas mixture in uniformly heated circular and tri‐lobe channels, numerical investigations are conducted and presented in the article. The adopted calculation models are validated by comparison between simulation results and experimental data. The local and total heat transfer characteristics are obtained, and new temperature correction factors for the circular and tri‐lobe channels are proposed and recommended. The simulation results obtained by using the SST k − ω turbulent model agree well with the experimental data, this kind of turbulent model matched with appropriate turbulent Prandtl number model could provide satisfactory temperature predictions for He‐Xe gas mixture in a wide range of temperature ratio Tw/Tb. The simplified Petukhov correlation with temperature correction can give conservative predictions with relative error less than 10%. For the tri‐lobe channel, due to the local restricted flow, the gas temperature at the corners is much higher than that at the concavities. The local wall temperature and local Nu number vary azimuthally, showing a large temperature gradient. The tri‐lobe channel has better heat transfer performance than that of the circular tube when the temperature ratio Tw/Tb is no more than 1.5, but it deteriorates significantly with the increasing temperature ratio. This article may contribute to the numerical simulation and thermal design of heat transfer components concerning He‐Xe gas mixture. [ABSTRACT FROM AUTHOR]

Summary: Accuracy and computational expenses are contradictory when performing simulation to hot plenum where multi‐dimensional phenomena voice loudly in sodium‐cooled fast reactor (SFR). A 2‐dimensional method was preferred in light of the cylindrical shape and drive force behind significant thermal stratification phenomena during transient case. One barrier related to buoyancy effect on turbulent flux transportation has been removed before. Asymmetric arrangement of components in circumferential direction makes another difficulty to reproduce actual flow field within hot plenum of pool‐type SFR. A series of special techniques for spatial discretization were implemented to restore geometry reality. Treatments about upper core structure, main heat exchanger and auxiliary heat exchanger were evaluated based on criterion of predicting well‐mixed inner plenum and stratification like outer plenum. Finally, optimal model came to an 8‐block partition scheme in which whole profiles of upper core structure and auxiliary heat exchanger were described. Validation was carried out through a normal state experiment on a water platform with reduced‐scale structure. Agreement of relative temperature distribution demonstrated ability of this model for tackling spatial layout in pool‐type SFR with the 2‐dimensional library. All real‐physics and efficient features indicate that this 2‐dimensional methodology is an excellent tool for engineering application when designing or evaluating a new SFR concept. [ABSTRACT FROM AUTHOR]

Summary: Space nuclear reactor power system (SNRPS), especially megawatt power level, is very attractive for the future civil and military space demands. The high‐temperature gas‐cooled reactor coupled with closed Brayton cycle (CBC) can achieve high‐power output for space applications. For the gas‐cooled SNRPS with CBC, dynamic models for all components, including reactor, turbine, compressor, alternator, ducting, pump, recuperator, gas cooler and heat pipe radiator, are developed in this article. Then, a transient system analysis code (SAC‐SPACE) is developed to perform the safety characteristics analysis of the SNRPS. In the full‐power steady‐state analysis, the maximum fuel temperature has a margin of 1323 K from the melting point. Moreover, the transient responses of the SNRPS under the mechanical failure of one Brayton loop (MFOBL1) and the reactivity insertion accident (RIA) are simulated and analyzed. During the MFOBL1, the maximum fuel temperature is 1275 K lower than the melting point, and finally, the reactor power stabilizes at 56.8% of the rated power. For the RIA, as long as the reactivity introduced into the core does not exceed 0.61 $, the SNRPS can reach a new steady‐state safely by itself without other protection strategies. It can be concluded that the SNRPS has good self‐stability due to the negative reactivity feedback of the reactor. This article may provide useful theoretical supports for the design and safety analysis of the gas‐cooled SNRPS with CBC. [ABSTRACT FROM AUTHOR]

Summary: The development of deep space exploration requires space nuclear reactor with higher power and better endurance. In order to meet the increasing high power requirements for space applications, an Open‐grid MEgawatt Gas‐cooled spAce nuclear reactor (OMEGA) is proposed in the paper, which is featured with low specific mass and high energy conversion coefficients. A set of models, including neutron kinetics model, reactivity feedback model and heat transfer model are established. Besides, the space nuclear system analysis code (System Analysis code of MEgawatt gas‐cooled space Reactor [SAMER]) is developed by finite volume method and staggered grid technique for the preliminary safety analysis of the OMEGA conceptual design. Furthermore, the calculation results provided by the SAMER code agree well with the computational fluid dynamics simulation, validating the reliability of the code. Moreover, the transient thermal‐hydraulic responses of the OMEGA under the unprotected reactivity insertion accident (URIA) and the unprotected loss of partial coolant flow accident (LOFA) are obtained and analyzed. The maximum fuel temperature after the URIA has a margin of 262 K from the fuel temperature limit value. The total power after the LOFA is about 68.7% of the rated power, meanwhile, the fuel radial temperature distribution tends to be more uniform with the fuel average temperature unchanged. It can be found that the OMEGA design shows satisfactory safety characteristic and reliability under the URIA and the LOFA scenarios. This paper is the stage summary of the conceptual design work of the OMEGA, and it may provide useful reference for megawatt space nuclear system design and thermal‐hydraulic analysis. [ABSTRACT FROM AUTHOR]

Summary: China Fusion Engineering Test Reactor (CFETR) has two operation phases: Phase I with Pf = 200 MW is to demonstrate the steady‐state operation; Phase II with Pf = 1 GW is to validate the DEMO technology. Helium Cooled Ceramic Breeder (HCCB) blanket is one blanket candidate for CFETR, and several research institutes have conducted extensive conceptual designs and comprehensive analysis works for Phase I HCCB blanket. However, because the operation phase of CFETR has transformed from Phase I to Phase II and the latest physical design parameters have been just determined, the design and analysis works for HCCB blanket should be initialized. In this work, first, the basic structure design of CFETR Phase II HCCB blanket is performed and a design scheme is proposed according to the design principles and requirements of blanket. Besides, the material selection of each blanket component has been determined. Second, based on the global neutronics model of CFETR Phase II, the neutron wall loadings (NWL) on different blanket modules are calculated. Afterwards, the radial structure layout optimizations of all poloidal blanket modules arranged in a 22.5° toroidal sector are conducted by NTCOC. As for Phase II blanket, which operates under the high fusion power condition, it should be noted that a dynamic feedback is incorporated into the original one‐way optimization method for solving the problems of inadequately optimized tritium breeding performance and insufficient tritium releasing ability. Finally, the radial structure layout parameters of all blanket modules in CFETR Phase II neutronics model are set as the obtained final optimal schemes. The calculation results show that the global TBR of CFETR Phase II is 1.182, which can well meet the tritium self‐sustaining requirement. This work can provide worthy guidance and reference for further conceptual design, comprehensive analyses and optimizations of the CFETR Phase II HCCB blanket. [ABSTRACT FROM AUTHOR]

• A sub-channel code which is capable of modelling the flow blockage has been developed and verified. • The results of flow blockage of the inner channel and out channel has been obtained. • The effect of the flow blockage on the heat split and DNBR has been discussed. A sub-channel code capable of modelling the flow blockage of the annular fuel has been developed and verified by the experiment. A fuel conduction model has been developed for the blockage to consider the effect of the blockage on the temperature distribution of the fuel. Blockage at different position has been modeled and the middle blockage is selected as the calculation object. Blockage of inner and outer channel with different blockage rate has been calculated and the results of fuel temperature distribution, the heat split, the mass flow rate and the DNBR distribution has been obtained. Based on the analysis of the results, the blockage of inner and outer channel will have different effects on mass flow and heat split of the fuel, and the blockage of inner channel has a greater impact on DNBR and fuel temperature than that of outer channel. [ABSTRACT FROM AUTHOR]

Summary: As one of the representatives of fourth generation nuclear reactors, molten salt reactor technology has become the research focus worldwide. Higher passive safety requirements have been put forward for the nuclear reactor systems. In this paper, a conceptual design of a passive residual heat removal system for the molten salt reactor is presented and the mathematical and physical models are established. Based on FORTRAN programming language, the passive residual heat removal system analysis code is developed and validated. The code is used to analyze the safety of molten salt reactor under accident conditions, including unprotected reactivity introduction accident, unprotected overcooling, unprotected loss of flow, opening failure of the frozen valve and the startup accident of partial heat pipes. The results show that the passive residual heat removal system of molten salt reactor can discharge fuel salt decay heat effectively within 200 hours and ensure that the molten salt reactor to operate safely under all accident conditions. Above all, the system has good passive safety performance. The proposed passive residual heat removal system can satisfy the safety requirement of molten salt reactor. The feasibility of the conceptual design of the heat pipe cooled passive residual heat removal system is theoretically verified. [ABSTRACT FROM AUTHOR]

• GeN-Foam has been extended to predict the occurrence of DNB. • CHF look-up table and post-CHF flow and heat transfer models have been developed. • OECD/NRC PSBT benchmark (Phase II) is used to preliminarily validate the new solver. • Fluid temperatures in each sub-channel resulted within ±10 °C with respect to experimental values. • The predicted DNB power and time of occurrence fell within a ±10 % error range of the experimental values for all cases considered in this paper. The prediction of departure from nucleate boiling (DNB) has always been a crucial aspect of thermal-hydraulic codes for the analysis of Light Water Reactors. In this paper, GeN-Foam, a multi-physics code developed based on OpenFOAM, has been enhanced to incorporate DNB prediction using a CHF look-up table as well as post-CHF flow and heat transfer models. To assess GeN-Foam's accuracy in modeling DNB conditions in typical pressurized water reactors (PWRs), a preliminary validation was conducted utilizing Phase II of the PSBT benchmark, including the steady-state fluid temperature (Test Series 1), steady-state DNB (Test Series 4), and transient DNB (Test Series 11T) benchmarks. This paper presents the closure correlations implemented in GeN-Foam, provides a description of the PSBT benchmarks, and includes a comparison between GeN-Foam's predictions and experimental data, as well as code-to-code verification with other participants. The results demonstrate that GeN-Foam exhibits good performance in simulating two-phase flow boiling conditions. Furthermore, GeN-Foam accurately predicts the DNB power and time of occurrence within a ±10 % error range compared to measured values, preliminarily indicating its effectiveness in predicting DNB occurrence and two-phase flow boiling phenomena. [ABSTRACT FROM AUTHOR]

The threshold displacement energy of Fe and Fe-Cr-Al alloys and the influence factors were investigated using molecular dynamic simulations. The mechanism of defect recombination was discussed, as was the evaluation of the recombination radius for Fe and Fe-Cr-Al. The simulation results show that Fe and Fe-Cr-Al have big differences in threshold displacement energies and directional dependencies at low temperatures, but the differences become less pronounced at high temperatures. The cause might be that the replacement collision sequence can effectively separate an interstitial from a vacancy in Fe-Cr-Al and greatly reduce threshold displacement energy at low temperatures. In addition, the energy transfer efficiency of primary knock-on Al atoms is low due to the difference in atomic mass, and the recombination effect for Cr is hindered by the formation of Fe-Cr 〈110〉 dumbbell. However, the effect of solute atoms on threshold displacement energy is insignificant at high temperatures, due to the increase in the recombination radius and a reduction in the effect of the aforementioned factors. [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]

MYRRHA is the Gen IV heavy liquid metal cooled fast reactor (LMFR), which is a pool-type reactor system developed and designed by the Belgian Nuclear Research Center SCK·CEN. The primary cooling system of the MYRRHA operates with molten lead-bismuth eutectic (LBE) as the working fluid, which offers several advantages. The equipment failures of fourth generation nuclear facilities will cause serious economic and environmental consequences, especially when flow blockage occurs, the threat to reactor safety is enormous. Rapidly detecting the occurrence of flow blockage and initiating shutdown procedures can effectively reduce the consequences of the accident. In this paper, a coupled sub-channel/CFD multi-scale platform is developed by implementing the sub-channel code SACOS-PB to the CFD code FLUENT based on the domine decomposition and explicit numerical scheme. The flow characteristics and temperature distribution in the upper plenum of a lead-cooled fast reactor(LFR) design based on MYRRHA reactor were investigated. The temperature at the core outlet increases 110K under the flow blockage condition, and the monitoring point for flow blockage is most suitable to be arranged in the center of the assembly and below the height of 0.6 m. The analysis results well demonstrated the 3D flow and local phenomena of the fluid in the upper plenum, which is significant to reactor design and reactor safety analysis of LFRs. [ABSTRACT FROM AUTHOR]

• A neutron diffusion model is developed based on OpenFOAM. • The model is validated by the benchmark. • The nuclear reactor neutronic thermal–hydraulic coupling code CorTAF-2.0 is developed. • The neutronic thermal–hydraulic coupling phenomena in the full core were carried out using CorTAF-2.0. Numerical reactor technology is the development direction of advanced nuclear energy system simulation, which has the ability of high-fidelity multi-physical field coupling calculation. The thermal–hydraulic and neutron physics fields in reactor core exhibit strong feedback relationships. Establishing the neutronic thermal–hydraulic coupling calculation model of the whole core "pin-by-pin" is the key technology for the development of numerical reactors. In this study, the neutron diffusion analysis model and cross-section database have been developed based on the open source CFD platform OpenFOAM. The three-dimensional neutron diffusion analysis model based on the finite volume method has been validated by benchmark problems. By integrating the neutron diffusion analysis model and cross-section database into nuclear reactor core three-dimensional thermal–hydraulic characteristics analysis code CorTAF, the strongly coupled neutronic thermal–hydraulic coupling analysis code CorTAF-2.0 has been developed. The neutronic thermal–hydraulic coupling phenomena in the full core of AP1000 have been simulated using CorTAF-2.0, yielding the sub-channel level parameters distribution of the whole core. This work offers valuable references for reactor full-core neutronic thermal–hydraulic coupling analysis. [ABSTRACT FROM AUTHOR]

• The visual entrainment experiments in the vertical tee branch are carried out, and the test data and videos are obtained. • The effects of branch Angle, air intake mode and working medium on entrainment are revealed. • The new OLE model and entrainment rate model are proposed, which can be used for nuclear reactor safety analysis. The tee branch structure is broadly used in the nuclear power systems, and liquid entrainment in the tee branch has been studied in depth. However, most of the existing research focuses on the vertical tee branch, while the study about the entrainment phenomenon in the inclined tee branch hasn't formed mature contents. This research aims to analyze the liquid entrainment in the inclined up tee branch. The liquid level measurement method of the Automatic Depressurization and Entrainment TEst Loop (ADETEL) was improved and supplementary experiments were conducted with single-end gas inlet of vertical tee branch. In addition, the test loop was modified to conduct visualization experiments on liquid entrainment experiments of the tee branch with a 45 degree angle. The experimental result reveals that the liquid entrainment phenomenon in the inclined tee branch is significantly different from that in the vertical tee branch. Based on the experimental phenomena and data, a new OLE (Onset of Liquid Entrainment) model is proposed which is well fitted with the existing data. Then, an entrainment rate model is proposed by nondimensionalizing the key parameters affecting the steady-state entrainment, with the error within ± 15% of the experimental data. [ABSTRACT FROM AUTHOR]

• Validation of THACS and OpenMC for SFR has been enhanced by simulating FFTF LOFWST Test #13. • A new modeling method multi-parallel channel plus local subchannel model is proposed for reactor core. • Important local phenomenon like inter wrapper flow, thermal stratification is analyzed by the specific model. In order to support collaborative efforts within international partnerships on the development of sodium-cooled fast reactor (SFR), the International Atomic Energy Agency (IAEA) proposes a coordinated research project (CRP) about the unprotected loss-of-flow-without-scram test (LOSWOS) performed at Fast Flux Test Facility (FFTF) in 2018, which could validate simulation tool and numerical models for SFR. As one of the participants, the Nuclear THermal-hydraulic Laboratory at Xi'an Jiaotong University (NuTHeL) participates with the monte carol code OpenMC and system thermal hydraulic code THACS. This paperdeal with the standalone neutronic modeling by OpenMC and thermal hydraulic modelling by THACS. For neutronic part, the axial homogenous model was modeled to analyze the reactivity feedback coefficients and steady radial power distribution. For thermal hydraulic part, the system model, including 2D CFD outlet plenum model, subchannel model and inter-wrapper flow model, was modeled to analyze both system response and local phenomena during the LOSWOS. In general, both neutronic results and thermal hydraulic results agreed well with other participants or experiments, the reactivity feedback of GEM could provide major negative feedback, which could be used as a novel passive shutdown device. However, the remaining discrepancies were the underestimated peak temperature of PIOTA at the begin of transients, it may be caused by unknow heat transfer coefficients around hexagon tube, which has no correlation and should be further investigated by experiments. In addition, when considering IWF model, the maximum temperature of the outlet for the active fuel region was 10 K larger than the subassemblie' outlet in the PIOTA, this phenomenon should be carefully treated in loss of flow accidents of SFR. [ABSTRACT FROM AUTHOR]

• In order to obtain the behavior of axial flow field evolution under the effect of intrinsic structure on fluid excitation, we focus on the fluid excitation of fuel rod with spacer grids in axial flow by three-dimensional numerical methods. • The effects of the clamping structures and mixing vanes are investigated separately, and their effects on the excitation force are analyzed by the simulation results. • The mechanism and causes of fluid excitation force generation are described. • The results of the excitation force and the frequency spectrum is analyzed. Grids, supporting fuel rods, consist of dimples and mixing vanes. These complex structures cause turbulent flow field with complex vortex. The vibration of fuel rod by the high-velocity coolant may lead to grid to rod fretting (GTRF). In order to obtain the behavior of the fretting wear, it is crucial to investigate the induction mechanism of fuel rod by the flow field evolvement. In this paper, the effects of two structures with different geometric were discussed by numerical simulation. The generation and evolution of the fluid forces are analyzed in detail. The results show that the fluid forces mainly originate from the pressure asymmetry. Moreover, the clamping structures play a dominant role in both excitation and fluctuation of the fluid forces. The study obtained the effect of intrinsic structure on fluid excitation, and it can provide a basis for the fluid excitation boundary of the fretting wear of fuel rods. [ABSTRACT FROM AUTHOR]

The strong coupling heat transfer between subassemblies and inter-wrapper gaps is closely related to the safety of the sodium-cooled fast reactor (SFR). Therefore, it is very important to carry out a detailed thermal hydraulic analysis of the core. In this paper, the inter-wrapper channel is modeled using the subchannel analysis method. A core thermal-hydraulic analysis code SACOS-Na (Subchannel Analysis Code Of Safety for SFR) is developed by using SIMPLE algorithm. For the subassembly module, a total of three experiments, including ORNL 19-pin experiment, TOSHIBA 37-pin experiment and KNS 37-pin experiment, are adopted to demonstrate the feasibility of the mathematical and physical models and numerical algorithms of the SACOS-Na as well as the rationality of the selected auxiliary models. For the inter-wrapper gap module, WEISS 2-subassembly experiment and CCTL-CFR 3-subassembly experiment are used to verify not only the simulation capability for backflow and strong inter-subassembly heat transfer, but also the heat structure model and the subchannel modeling approach on inter-wrapper gaps. In addition, the prediction accuracy of SACOS-Na has reached the advanced international level by comparing with the worldwide well-known core thermal hydraulic analysis codes. • The subchannel modeling method is adopted to build the mesh for inter-wrapper channels. • The SIMPLE algorithm is used to solve the problems of the reverse flow and inter-subassembly heat transfer occurred in SFRs. • The universal core thermal-hydraulic analysis code SACOS-Na for SFRs is developed. • The accuracy and reliability of the SACOS-Na are discussed from all aspects. [ABSTRACT FROM AUTHOR]

• A novel and efficient mesh grids generation and data mapping methods is proposed. • The coupling calculation of neutronics code OpenMC, fuel performance analysis code BEEs-Plates, and nuclear reactor system analysis code NUSAC has been implemented based on the open-source MOOSE platform. • The multiphysics coupling analysis is realized for dispersion plate-type fuel assembly. The nuclear reactor is a multi-physics and multiscale coupling system. The conditions of high temperature and high neutron flux in the reactor pose a significant challenge to the thermodynamic performance of the fuel. To evaluate the plate-type fuel's performance, a coupling calculation of neutron physics code OpenMC, fuel performance analysis code BEEs-Plates, and nuclear reactor system analysis code NUSAC has been implemented based on the open-source MOOSE platform. A novel and efficient method for mesh grids generation and data mapping between external codes and MOOSE was proposed to ensure accurate data exchange. This method can simplify the mesh grids generation process of external codes and accurately transfer data across different mesh grids. Finally, the coupling calculation of the JRR-3 M research reactor fuel assembly was carried out based on the above coupling code, and the neutronics parameters, thermal–hydraulic parameters, and fuel performance-related parameters of the fuel assembly were analyzed. The calculation results show that the maximum fuel burnup will reach 16.91% FIMA after 240 days. With the increase of fuel burnup, the peak power of the fuel assembly is significantly shifted towards both ends of the height direction. The VonMises stress, volume stress, creep strain, and displacement of the fuel will be considerable. Finally, the impact of the coupling calculation on the accuracy of the computational efficiency is preliminarily investigated. [ABSTRACT FROM AUTHOR]

• Current models of HTHPs are reviewed. • Heat transfer limits and failure of HTHPs are summarized. • Design and optimization of heat pipes are discussed. • Potential applications and challenges of HTHPs are concluded. Adopting liquid metals as the working fluid, high temperature heat pipes (HTHPs) can operate at a high temperature above 750 K and have extremely high heat transfer capacity, while the most significant weakness of HTHPs is frozen startup and heat transfer limits that lead to the failure of HTHPs. Previous studies on the theoretical models, design-optimization methods, and applications are reviewed and discussed critically. The operation stage of HTHPs is categorized into transient operation stage (lumped model, network thermodynamic model, and heat and flow model), frozen startup (fusion and vapor transition), heat transfer limits and failure. Basic assumptions and limitations of heat pipe models are outlined, and theoretical comparisons are drawn with respect to the applications and results. A review of the design procedure and optimization methods with constraints in HTHPs is presented, indicating that a design-optimization method involving the various wick structures, frozen startup process, and associated thermal performance variables has not yet been established. This review establishes that HTHPs offer viable potential and spread value for optimization and integration into nuclear reactor systems, space exploration, and renewable energy systems, while challenges for the HTHPs lie on (1) heat pipe models need to be improved to consider practical factors including volumetric filling ratio, composite wick, etc.; (2) heat transfer limits and failure mechanism and models verification; (3) incorporation of heat pipe models into optimization algorithm with acceptable efficiency; (4) heat transfer power enhancement and special-shaped HTHPs in potential applications. [ABSTRACT FROM AUTHOR]

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