bluuebunny/crossref_metadata_2025_split huggingface dataset:
Neutronics Research on Annular Axial-Non-Uniform Minor Actinides Transmutation Targets in Commercial Pressurized Water Reactor
Authors: Ye Bin , Ala Ayodeji A. 
Published: November 2024

Abstract
This paper introduces the thermal-fast neutron conversion material 6LiD to improve the transmutation efficiency of MAs in commercial PWR, especially the fission rate, and investigate the feasibility of using annular axial-non-uniformly transmutation targets to flatten the axial power distribution of the core and optimize the temperature reactivity effect of the moderator at high boron concentration. In this research work, four annular axial-non-uniform transmutation targets are designed and they have the same geometry structure, but axially have different MA concentration in the transmutation material which composed of MA, 6LiD and UO2. The annular transmutation targets are axially three, five and seven segments respectively, and MA concentration in the transmutation material gradually decreases from the middle to both ends. The transmutation performance of these targets is investigated while its parameters change, such as the thickness of the annular transmutation rod and the composition of the transmutation material. The results show that, compared with axially uniform rods, the annular axial non-uniform transmutation targets have high transmutation efficiency, especially the fission rate. The transmutation effect is the best while the mass ratio of 6LiD to MAs in the transmutation material is 1:9; MA fission rate of these transmutation targets can reach up to 5.05%. At the same time, these targets have good power flattening effect, especially axial power flatten results that they could reduce the axial power peak factor of the core from 1.778 to 1.388. Meanwhile, they can also optimize the temperature effect of the moderator well at high boron concentrations.
An Improved Multiphase Lattice Boltzmann Method Scheme for Large Density Ratio Simulation in Heat Pipe
Authors: Quan Lie , Ma Yugao , Liu Xiao , Huang Shanfang , Wang Kan 
Published: November 2024

Abstract
Heat pipe reactor is a cutting-edge energy power technology for future deep space and deep sea exploration, its current bottleneck is the development of high performance heat pipe, and the wick structure is the key factor affecting the performance of heat pipe. To promote research on flow and heat transfer mechanism inside heat pipe, an improved scheme of Multiphase LBM is introduced, which not only eliminates error caused by inconsistency of discretization scheme of pressure term when introducing non-ideal gas equation of state, but also the cubic term error of velocity field in the recovered Navier-Stokes equation through LBM. With this improved scheme introduced, results of better accuracy can be obtained, so as its application into permeability prediction of porous media. An ideal Kelvin structure is constructed according to the production process and geometric parameters. The lattice Boltzmann method was used to directly simulate the three-dimensional flow inside Kelvin structured metal foam on the pore scale, and the permeability are measured through numerical experiments based on Darcy’s Law.
When solving the multiphase flow problems such as gas-liquid flow with multiphase Lattice Boltzmann method (LBM), the gradient of the macroscopic quantity is required to calculate the interaction forcing term between the neighboring fluid particles. It is found that unphysical numerical oscillations would occur in the vicinity of the phase interface. Through analysis and test, this numerical oscillation is found to be related to the discretization scheme of gradient operator between the phase interface, and will lead to numerical instability, which limits the simulation ability of LBM in dealing with the two-phase interface problem with high density ratio. Simulations of flat interface problem and liquid droplet system are made to confirm better performance of this optimized forcing scheme, and results of higher resolution of the flow field was obtained, which are in good agreement with analytical solutions and experimental results.
TRIPOLI-4® Monte Carlo Code Verification and Validation Using T4G Tool
Authors: Lee Yi-Kang , Hugot François-Xavier 
Published: November 2024

Abstract
TRIPOLI-4® is a general-purpose Monte Carlo radiation transport code developed at CEA Paris-Saclay center. It uses continuous-energy nuclear data to simulate neutron, photon, electron and positron transport in fields like radiation protection and shielding, fission reactor physics, nuclear criticality safety, and nuclear instrumentation. The verification and validation assessments of TRIPOLI-4® input and output (I/O) data are one of the fundamental processes to assume the accuracy and credibility of numerical studies for its nuclear engineering applications.
T4G is an interactive tool for the TRIPOLI-4® I/O data display. It is a graphics viewer displaying 2D and 3D plots of geometry, materials of volume-cells, radiation source, collision sites projection of particle transport in space and energy, particle trajectories and leakage sites. It also provides iso-importance curves for variance reduction study, tally results display including iso-dose and iso-reaction rates curves, and volume-cell numerical values. T4G tool relies upon the same geometry module as TRIPOLI-4® code and therefore displays exactly the geometry that TRIPOLI-4® uses for neutron and photon tracking.
In this paper, we first introduce the TRIPOLI-4® code and its verification and validation (V&V) study. Secondly, we present the basic features of T4G display tool. Then we demonstrate and discuss advanced application cases summarizing the real practices of T4G tool for verifying and studying of TRIPOLI-4® I/O data in diverse fields of nuclear engineering.
Experimental Study on the Influence of Spray Characteristics and Temperature Control on the Removal Efficiency of Iodine Aerosols
Authors: Tang Jiaxuan , Yang Yang , Zhang Luteng , Li Jialong , Pan Liangming , Liu Zhuo , Gao Li , Yuan Yidan 
Published: November 2024

Abstract
Radioactive aerosol may be released into the gas and retained in the containment under severe nuclear accidents conditions. Spray system removal is one of the main ways to remove aerosols. Spray droplets and overpressure of primary loop in severe accidents create a complicated high temperature and high humidity environment. A significant portion of radioactive aerosols in nuclear accidents are soluble. These soluble radioactive aerosols (CsOH/CsI) in the containment of the high humidity environment will occur significant hygroscopic growth behavior and may dissolve in droplets. Moreover, if the spray solution is alkaline, changes in particle density, size, shape, and chemical composition of iodine aerosols may occur, significantly influencing the dynamic behavior of aerosol agglomeration, diffusion and deposition, and consequently, the aerosol removal efficiency of the spray system. While extensive research has been conducted on the spray removal of insoluble aerosols (TiO2\SiC), studies on typical soluble iodine aerosol spray removal experiments are largely lacking.
In this paper, the experimental study of single nozzle with different spray characteristics on the removal efficiency of iodine aerosol was carried out. This study provides insights into the effects of different spray pressures, flow rates, and pH values on the removal efficiency and influencing patterns of soluble iodine aerosols under various initial particle sizes and different aerosol compositions. Results indicate that, Alkaline spray water with different pH values had no significant effect on the removal efficiency of iodine aerosol. The aerosol removal coefficient in the 0.2μm and 0.5μm particle size ranges of iodine aerosols is the smallest, which is 0.0007s−1–0.0012s−1. Increasing the proportion of soluble aerosols in multi-component aerosols will increase the removal rate of spray aerosols.
This study can provide reference for fundamental understanding of the aerosol removal mechanism of soluble iodine aerosols by containment spray droplets and improving the containment spray removal models.
Generative Design of a Gas-Cooled Micro Reactor Based on Multi-Physics Analysis
Authors: Han Wenbin , Lu Qi , Deng Jian , Huang Shanfang 
Published: November 2024

Abstract
Additive manufacturing (AM) is a “game-changing” digital fabrication technology that can fundamentally alter the way products are designed and manufactured. In nuclear engineering, the application of additive manufacturing can enable advanced reactor designs with better performance, lower cost, and quicker deployment. To exploit the maximum potential of additive manufacturing, design for additive manufacturing (DfAM) is needed, DfAM can leverage the complexity freedom offered by additive manufacturing technology to break traditional design constraints and achieve enhanced performance. Generative design, one of the key technologies in DfAM, is an iterative exploration method using AI-driven algorithms to generate a series of design solutions that meet specified constraints.
In this work, a generative design platform for micro-reactors is developed based on DASSULT SIMULIA Isight. To simulate and analyze the multi-physics performance of the reactor design, various multi-physics simulation codes are integrated on this generative design platform, including RMC for neutron physics analysis, ANSYS Fluent for computational fluid dynamics analysis, and ANSYS Mechanical for structural mechanics analysis. A multi-objective optimization method is used to explore geometric parameters and fuel configurations to achieve better neutronics, thermal-fluidics, and structural mechanics performance.
The generative design solution is a 4 MW helium-cooled nuclear reactor which includes 3D-printed silicon carbide (SiC) matrix embedded with tri-structural isotropic (TRISO) particle fuel. Helium at 3 MPa is chosen as the coolant, and the cross-shaped coolant channels are distributed within the matrix to remove the heat generated by fission. According to the multiphysics analysis, the reactor has a initial keff of 1.0571, power peaking factor FQ of 1.50, and operating lifetime of 1200 EFPD (Effective Full Power Days), with a maximum temperature of 1331 K and a peak thermal stress of 141.3 MPa.
The generative design method and platform provide a systematic and efficient approach for the design of additively manufactured micro-reactors, and the generative design solution demonstrates the feasibility and effectiveness of the proposed method.
Research on Dynamic Rod Measurement Method Based on Torch Software Package for PWR
Authors: Cai Yun , Li TianYa , Guo Rui , Liu Tongxian , Li Qing , Yang LingFang , Zhao Wenbo 
Published: November 2024

Abstract
Dynamic rod measurement method proposed by Westinghouse and Electric Power Research Center is an effective control rod worth measurement method, which has been applied widely. The main feature of dynamic rod measurement method includes the high accuracy and the short process of the corresponding rod worth measurement test. Torch software package developed by Nuclear Power institute of China can be used to analyze the neutron physics of PWR. Corca-K in the Torch software package is the tool that is used to solve the three-dimensional space-time neutron kinetics equations. A tool based on dynamic rod measurement method and Corca-K was developed, and verified by the control rod worth measurement test of Unit #1 of FangJiaShan and Unit #4 of FuQing power plant. The results showed that the developed tool was accurate, which the relative error of the different control rods between the calculated control rod value and the measured control rod value was less than 10%. The results showed that the developed tool could be applied to the measurement of the control rod value.
Research on Multi-Scale Coupling Program Based on Lead Bismuth Fast Reactor
Authors: Zhang Yixin , Wang Chenglong , Zhang Jiaxin , Tian Wenxi , Qiu S. Z. , Su G. H. 
Published: November 2024

Abstract
With the increasing emphasis on the refinement of reactor thermal hydraulic calculations, research on multi-scale coupled models of nuclear reactor thermal hydraulic has also developed. This article utilizes the self-developed lead bismuth system analysis program SACLER, combined with the mature commercial CFD software FLUENT, to develop a coupled program suitable for 1D/3D calculations of lead bismuth fast reactors. In order to ensure the accuracy of the developed coupling program, the calculation function of the developed coupling program was first verified through a simple circular tube case. The coupling program matches well with the data calculated by SACLER and FLUENT respectively, with temperature and flow rate errors within 10%. Subsequently, the response of the coupling program to the thermal hydraulic phenomena of the lead bismuth reactor under transient operating conditions was explored, and the transient experiment on the NACIE facility was simulated. The calculation results showed that the coupling program was in good agreement with the experimental results and could effectively simulate the changes in transient thermal hydraulic parameters.
Numerical Simulation of Single/Two-Phase Flow in a Stratified Porous Bed
Authors: Bouloudenine Aimad , Li Liangxing , Xiang Zutao , Shi Shang , Abu Bakar Muhammad 
Published: November 2024

Abstract
This study investigates the coolability of debris beds formed from Fuel-Coolant Interactions (FCI) within nuclear reactors, a critical safety concern when cooling systems fail, leading to significant core melting in light-water reactors. The research focuses on understanding the flow dynamics within two-layer stratified porous beds, which develop from the mixture of corium and coolant, particularly around the interface between layers. Using ANSYS Fluent software, numerical simulations were conducted to estimate velocity profiles and pressure drops across a stratified porous structure. The Euler-Euler two-fluid model was employed alongside the k-ω-based Shear Stress Transport (SST) turbulence model to separately model turbulence for each phase. The results reveal that lateral flow dynamics are significant as fluids traverse through the porous media, affecting pressure drops across layers with varying permeabilities. Specifically, the pressure drops across the lower-permeability layer decrease while that across the higher-permeability layer increases, highlighting the influence of lateral flows between layers of different particle sizes due to variations in porosity. This lateral flow is most likely formed during the initial stages of fluid movement through the bed. The study underscores the importance of understanding these flow characteristics for both single/two-phase flow to enhance the safety and design of nuclear reactors by improving debris bed coolability predictions.
Prediction of Critical Heat Flux of Wire-Wrapped Rod Bundle Based on Artificial Neural Network
Authors: Zhang Wei , Xiao Yao , Yu Lijun , Gu Hanyang 
Published: November 2024

Abstract
Accurate prediction for the critical heat flux of rod bundle assemblies is important for reactor safety assessment. However, the existing prediction methods fail to unify their generalization and accuracy. The theoretical prediction models are unable to provide a satisfactory estimation. while the empirical methods that have been proposed for several decades could give results precisely, their suitable conditions are strictly limited, causing the endless development of different correlations for every specific rod bundle design. An exploration of new methods has been undertaken to provide a better method for critical heat flux prediction based on an artificial neural network since neural networks have gained recent attention for their outstanding performance in predicting complex problems. This study discusses the prediction performance of a traditional polynomial fitting method, a neural network method, and a hybrid method combined with empirical correlations and a fully connected neural network regression model. The analysis concludes that when artificial neural networks replace empirical models in prediction tasks, achieving equivalent prediction accuracy is easily accomplished; however, there is a need for improvement in error characteristics. Besides, the hybrid method was validated as a superior approach, demonstrating its effectiveness in enhancing prediction performance across various input variable regions. The ratio of predicted values to true values is well-maintained, indicating that this proposed method yields improved error characteristics, and consequently, more accurate critical heat flux estimations. The research will help improve the feasibility of the application of machine learning algorithms such as neural networks in the nuclear energy industry.
Study on Flow Fluctuation Attenuation in the L-Shaped Pipe With Spring Vibrator
Authors: Wang Sipeng , Chen Jin , Kang Zehang , Yue Zhongcheng , Zhuang Kun 
Published: November 2024

Abstract
In view of the flow instability phenomenon that often occurs in passive residual heat removal system (PRHRS), the idea of using spring vibrator structure to attenuate flow and pressure fluctuations is proposed. Based on the finite element method, the fluid-solid coupling analysis is carried out in a typical L-shaped pipe which contains a spring vibrator structure. The results confirm the fluctuation attenuation effect of the spring vibrator structure. The simulation results shows that flow fluctuation amplitude at the outlet reduce, and the peak pressure which is the most threatening to thin-walled structures reduces. Based on the influence of relevant parameters on the attenuation coefficient, an optimized spring vibrator structure is obtained by changing the values of relevant parameters. Comparing the outlet flow fluctuation before and after the optimization of the spring vibrator structure, the flow fluctuation amplitude at the outlet is significantly reduced. The pressure fluctuation is greatly weakened compared with that before the optimization. The simulation results prove that the optimized spring vibrator has excellent fluctuation attenuation performance.
Effect of Thermal Cycling on the Microstructure of Nuclear Grade Dissimilar Metal Weld
Authors: Luo Wenwei , Hu Mengjia , Luo Xiaowei , Shi Li , Wu Xinxin 
Published: November 2024

Abstract
As the crucial structure in the pressure boundary of the steam generator in the high-temperature gas-cooled reactor (HTGR), it is essential to investigate the microstructure of T22/800H dissimilar metal weld (DMW) under thermal cycling. This is especially critical in the context of rapid temperature fluctuations under extreme conditions, which are likely to inflict irreparable damage on the material’s microstructure, thus posing a severe threat to the service life of the DMW. T22 and 800H base metal (BM) were welded by tungsten inert gas (TIG) welding, filling with UNS N06082 inserts. To investigate the impact of thermal cycling under extreme conditions on the microstructure of DWM, and to accelerate the evolution of microscopic phenomena, simulated thermal cycling tests were carried out at 500°C with a holding time of 1 hour per cycle and 300 cycles at two ramp-up/down rates of 3°C/min and 8°C/min between room temperature and 500°C, respectively, much more serious than the temperature evolution in start-ups and shutdowns during the operation of HTGR. Characterization of the microstructure of three typical failure-prone areas of T22/800H DMWs, namely the T22 base material (BM), the T22 heat-affected zone (HAZ) and the T22/WM fusion zone by optical microscopy (OM), scanning electron microscopy (SEM) and energy dispersive spectroscopy (EDS), compared with that of the as-welded.
It is obtained that limited cavities formed at the interface of T22 BM and the weld fusion zone (FZ), due to the prominent difference of thermal expansion, which contributes to the creep and fatigue, especially in the thermal cycling. Some cavities were also found in the fine-grained heat-affected zone (FGHAZ) of T22, ranging from 100 to 700nm in size. It is predicted that the cavities in FGHAZ were resulted from stress relief without preheat and post-weld heat treatment during the manufacturing of T22/800H DMWs. In addition, a mount of cavities, less than 100nm in size, were observed in the T22 BM during the thermal cycling test at 8°C/min; meanwhile, there were also some cavities, ranging from 100 to 300nm, found in the coarse-grained heat-affected zone (CGHAZ). It is supposed that the differences of thermal-physical properties, such as thermal expansion and hardness, make sense for the small cavities under thermal cycling. Sharp notches resulted from oxidization and thermal stress were found at the interfaces between T22 BM and weld metal, except for specimens without simulated thermal cycling.
Coarse Mesh Generation for CMFD Acceleration in SARAX-LAVENDER Code
Authors: Xu Haoxiang , Zheng Youqi , Xiao Bowen , Wu Hongchun 
Published: November 2024

Abstract
A coarse mesh generation algorithm is implemented in SARAX-LAVENDER code based on mesh merging, aiming at providing the foundation for a CMFD acceleration scheme. Originated from Navigation Mesh construction method in Recast code, this algorithm is slightly altered to improve performance on a reactor geometry. The algorithm can handle regular triangular node in structured geometry as well as Delaunay triangle mesh in arbitrary unstructured geometry. Such an algorithm holds several advantages as follows: (1) No extra effort is required during regular triangle or Delaunay triangle mesh generation procedure. (2) The coarse mesh generated by triangle mesh consists of only convex polygons. (3) The boundary surfaces between coarse mesh can be exactly preserved by existing triangular node. (4) The maximum edge number of generated polygons could be easily adjusted to adapt different geometry. When set to 3, the algorithm could transfer triangle mesh directly into coarse mesh format without mesh merging, making it easy to implement a two-level spatial CMFD acceleration scheme. (5) The algorithm naturally adapts the geometry after domain decomposition, which is required for parallel computing. Coarse mesh number ranges from one sixth to one third of original triangle mesh number in hexagonal, rectangle or arbitrary unstructured geometry. Similar reduction could be achieved in a MPI parallel computing condition with CPU number from 2 up to 64.
Sigma Embrittlement Evaluation Test for Dissimilar Welding Between F6NM and FXM-19
Authors: Osafune Ryoji , Tanabe Yoshihiro , Yagi Daisuke , Koike Daisuke 
Published: November 2024

Abstract
In dissimilar welds between martensitic stainless steel F6NM and nitrogen-strengthened austenitic stainless steel FXM-19, type 209 austenitic welding materials is used to match the mechanical properties and chemical composition of FXM-19. The welds of F6NM must be subjected to Post-Weld Heat Treatment (PWHT) to restore ductility and toughness. Therefore, sigma embrittlement is a concern for ER209 buttering welds as they are subjected to PWHT. In this research, the mechanical properties and microstructure of dissimilar welds of F6NM+FXM-19 were investigated.
There was no significant difference in the volume fraction of sigma phase between the buttering weld metal (PWHT) and the groove weld metal (As weld) for each welding process. Therefore, sigma phase transformation by PWHT did not occur. Submerged Arc Welding (SAW) and Gas Metal Arc Welding with 100%Ar gas shield (GMAW 100%Ar) were found to have good mechanical properties. In Gas Tungsten Arc Welding (GTAW), the impact value was below the required value in the Charpy impact test of the buttering weld metal. SAW and GMAW 100%Ar were found to be applicable welding processes for this dissimilar weld joint.