bluuebunny/crossref_metadata_2025_split huggingface dataset:
Numeral Research on Thermal-Hydraulic Characteristics of Pebble Bed in WCCB Blanket Under In-Box LOCA
Authors: Bai Hanlin , Cheng Jie , Han Shuaiyu , Wu Di , Wang Jianjun 
Published: November 2024

Abstract
After the breakage of the coolant pipeline in Water Cooled ceramic Breeder blanket (WCCB), the coolant at high pressure inside the pipeline will be sprayed into the pebble bed environment, and the specific behavioral characteristics of this process are not clear. In order to investigate the flow characteristics of the coolant in the pebble bed after the rupture occurs, numerical simulation of this physical process is carried out based on the three-dimensional computational fluid dynamics (CFD) method. The flash phase transition process is described by the VOF model and the buckling flash model. The simulation results show that, after the In-box LOCA occurs, the coolant enters the proliferating spherical bed and rapidly vaporizes into water vapor by flash evaporation. Due to the geometrical characteristics of the blanket structure, the water vapor diffuses inside the pebble bed and then gradually diffuses to both sides and contacts the underside of the pebble bed; in addition to this, helium will flow out of the cladding along the helium pipeline during the In-box LOCA process; The In-box LOCA process can be initially divided into 2 stages. Stage 1: water vapor rapidly diffuses to contact the closer wall of the pebble bed at the beginning of the breach, which has a longer duration and a faster rate of pressure rise. Stage 2: water vapor diffuses mainly to the farther side of the wall, and because the gas flows out along the helium pipeline, this stage will last longer and the pressure rises slowly and is relatively stable.
Investigation on Applicability of Dynamic Mode Decomposition in Burnup Analysis
Authors: Xue Pengchao , Zhang Qian , Zhao Qiang , Zhang Yunfei 
Published: November 2024

Abstract
In burnup analysis of a reactor, neutron transport calculation and depletion calculation are carried out alternately, bringing a serious computational burden to large-scale problem calculation. In this study, we used a data-driven strategy based on dynamic mode decomposition to extract the modes and characteristics of nuclide density to construct a prediction expression for the evolution of nuclide density. Reducing the burden of burnup calculation can be achieved through direct prediction of the nuclide density in a certain time point in the future based on the dynamic mode decomposition algorithm[1]. Firstly, six snapshots of different size are selected to test the relationship between the size of snapshot and prediction accuracy. The results indicate that when the time scale of the snapshots is the same as the time scale of the period to be predicted, the dynamic mode decomposition algorithm can extract a more comprehensive spatiotemporal evolution model. Secondly, we selected appropriate-sized snapshots to test the impact of different truncated rank on prediction accuracy. The results show that the optimal truncated rank for the assembly-level problems is 11. Thirdly, we explored the effect of the spatial field information on the results of prediction. The results indicate that using snapshots containing spatial information for prediction does not effectively improve accuracy and still incurs a high computational burden. Finally, the applicability of the dynamic mode decomposition algorithm to different fuel assemblies and the computational efficiency of the DMD algorithm are tested. The results show that the prediction deviation of nuclide density for different assemblies is around 1%, while saving 99.87% of the burnup calculation time. The result proves that the dynamic mode decomposition algorithm can efficiently discover the hidden dynamics in burnup analysis and has a considerable application prospect in large-scale burnup analysis.
Molecular Simulation of I2-H2O Diffusion in Different Pore Structures of Activated Carbon Fiber
Authors: Li Xin , Liu Jingguo , Shuang Yingzhong , Liu Zhixin , Li Jian , Yu Jie , Li Yongguo , Liu Liye 
Published: November 2024

Abstract
Activated carbon fiber (ACF) is a new, efficient, and multifunctional adsorption material developed in the 1970s. In order to develop an iodine adsorbent of ACF, the diffusion effect of iodine in the pores of ACF was studied. Based on molecular dynamics, and Monte Carlo methods, a pore structure model of ACF was constructed based on carbon nanotubes. The accuracy of the model was verified through diffusion experiments and comparison with simulation results. And then, according to the basic structural unit of the constructed ACF pores, the mean square displacement and radial distribution functions of I2 and H2O under different pore sizes, functional groups and temperatures were studied, and the interaction between I2-H2O was obtained. The results show that for different diffusion components at the same temperature, the diffusion rate shows I2 &gt; CH3I &gt; H2O with the increase of time step. The diffusion rate of I2 is positively correlated with temperature and pore size. For different relative humidity, the diffusion rate is 40% &gt; 80% &gt; 10% with the increase of time steps. This study provides a theoretical basis for achieving the goal of iodine removal of activated carbon fiber.
Study on the Effective Diffusion Coefficient of Inert Gases in Porous Graphite Materials
Authors: Nie Rui , Wang Yu , Zhou Ziling , Zhang Weihua , Xie Feng , Li Hong , Wei Liqiang , Fan Zhixiang , Fu Jia , Fan Qunchao 
Published: November 2024

Abstract
The high-temperature gas-cooled reactor (HTGR), is globally recognized as one of the reactor types embodying the safety features of the fourth generation, with the advantages of safety, high thermal efficiency in power generation, a small emergency planning zone, and the ability to use high-temperature helium for hydrogen production. HTGRs utilize a significant amount of graphite materials, including fuel graphite matrices, graphite moderator material, and graphite reflector layers and so on. Fuel containing fissile materials produce inert gas fission products such as Kr and Xe, while isotopes like Ar-40 in the pores of graphite materials can become activated, forming inert gases such as Ar-41. The fuel element matrix graphite is the necessary pathway for these inert gas nuclides to enter the primary coolant circuit. However, current research rarely provides a comprehensive analysis of the diffusion behavior of inert gases in porous graphite materials. This study establishes a computational model and methodology for determining the effective diffusion coefficients of inert gas nuclides in porous material graphite. It quantitatively calculated the cumulative release fractions and release ratios of Ar, Kr, Xe, in both open and closed pores of the fuel element matrix graphite and elucidated the impact and patterns of factors such as pressure, temperature, and half-life on the diffusion behavior of inert gases in porous material graphite. This research provides valuable insights for future studies on the diffusion behavior of Ar, Kr, and Xe in graphite.
Conceptual Design of High-Power Nuclear Power System on the Surface of Mars
Authors: Han Yansong 
Published: November 2024

Abstract
Mars has always been an important target of deep space exploration, and it is expected to achieve the goal of manned landing around 2050. For manned Mars missions, a reliable and powerful power system is essential. Because of its simple, reliable, safe and high energy conversion efficiency, sCO2 Brayton cycle with heat pipe reactor has become a promising high-power nuclear power scheme on the surface of Mars. In this paper, the concept design of Mars surface megawatt high-power nuclear power supply system based on sCO2 Brayton cycle with heat pipe reactor was carried out. In order to explore the native application potential of Martian surface atmosphere, two working fluids (pure CO2 and the CO2+N2+Ar mixture of Martian surface atmosphere) were compared. The thermodynamic properties of working fluid in supercritical state are complex. Therefore, it is very important to use appropriate thermodynamic property calculation tools. In this paper, the commonly used calculation tools for the thermodynamic properties of working fluid were verified to select the appropriate tools. Then, the concept of Mars surface megawatt high-power nuclear power system was constructed, and the thermodynamics and mass models were completed. On this basis, the operation characteristics of the system were analyzed, and the application potential of Martian atmospheric mixture was preliminarily evaluated. Finally, the system was comprehensively optimized to determine the optimal parameter setting, finishing the conceptual design of the system and the exploration of application value of Martian surface atmosphere.
Study on Long-Life Minor Actinide Temporary Storage Scheme Based on High-Level Liquid Waste Partitioning Strategy
Authors: Meng Xiantao , Wei Meng 
Published: November 2024

Abstract
At present, transmutation and fast reactor fuel cycle cannot be implemented in engineering, minor actinides obtained by high-level liquid waste partitioning may be managed by the concept of “conditioning - temporary storage - retrieval - reuse”. In this paper, Am/Cm stripping stream of TRPO process is analyzed as example of this concepts. The physical properties (composition, mass, radioactivity, heat power) of the americium-curium mixed oxide, and the variation of these properties over time have been investigated. Although the mass ratio of Am and Cm in the Am/Cm stripping stream obtained by TRPO process is very low (less than 5%), the Ln/An separation may not be necessary under the concept of “conditioning - temporary storage - retrieval - reuse”, because during the temporary storage of americium-curium oxides, U, Pu, Np will be produced through alpha decay of Am and Cm. The rare earth elements and transition metals contained in the Am/Cm stream are the main contributors of β/γ radioactivity and short-term heat in the mixed oxides, but the β/γ radioactivity and short-term heat release will decay rapidly, leaving only a radioactivity and long-term heat of Am and Cm.
Dimensionless Analysis of Aerosol Dynamic Behaviors in the Light Water NPP Severe Accidents
Authors: Wang Jinghong , Peng Wei , Yu Suyuan 
Published: November 2024

Abstract
The aerosols dispersed inside the containment space undergo a variety of complex dynamic behaviors such as convection, sedimentation, thermophoresis, diffusiophoresis, and Brownian coagulation after a severe accident in a modern light water nuclear power plant (NPP). These aerosol behaviors occur simultaneously and compete with each other. The relative significance of the aerosol behaviors varies greatly under different environmental conditions. In order to study the significance of convective transport, gravitational sedimentation, thermophoresis, diffusiophoresis, brownian coagulation under different environmental conditions after a severe accident, a population balance equation for aerosol dynamic behaviors after a severe accident was constructed in the present work. A dimensionless analysis was carried out on the population balance equation, and the Peclet number and the Damköhler number, were introduced to describe the relationship between the aerosol behaviors after a severe accident. The variation rules of the dimensionless parameters were obtained under various environmental conditions such as temperature, pressure, particle sizes, and particle number concentration. The results show that, under typical severe accident conditions, the changes of temperature and pressure have significant effects on the diffusiophoresis process. The change of aerosol number concentration mainly affects the aerosol diffusion and coagulation rate. While the aerosol particle size affects most aerosol dynamic processes, and the gravity sedimentation process has the highest sensitivity to the change of particle size. The results of this study are helpful to understand the dominant mechanisms of aerosol behavior after severe accidents and to understand the spatial-temporal evolution of aerosols after severe accidents.
Surrogate Modeling Advancements in Nuclear Fuel Engineering: Bridging the Gap Between Accuracy and Efficiency
Authors: Cancemi Salvatore Angelo , Angelucci Michela , Lo Frano Rosa , Paci Sandro 
Published: November 2024

Abstract
In nuclear reactor fuel engineering, the phenomena of Pellet-Cladding Interaction (PCI) and Pellet-Cladding Mechanical Interaction (PCMI) present significant challenges. These issues affect the design and safety of NPPs, mainly due to factors like fission gas release and fuel swelling. Accurately modeling these interactions is complex, as they are complexly linked with the fuel’s neutronic and thermal responses. Traditionally, the simulation of these interactions, essential for nuclear safety and risk assessment, has been complex and resource intensive. This study focuses on developing innovative surrogate models to enhance computational efficiency in nuclear fuel code analysis. A surrogate model based on a neural network approach is a data-driven computational model that approximates the behavior or output of complex, time-consuming, or resource-intensive simulations. This study specifically employs Artificial Neural Networks (ANNs) and statistical algorithms, aiming to reduce high computational cost of traditional approaches. A validated synthetic dataset, representing thermal analysis under steady-state conditions, is used to train the machine learning model. The dataset specifically focuses on Cladding Temperature. The study investigates eight different Test-Case. The surrogate model, trained on only 20% of the dataset, can predicts the entire time series of temperature using ARIMA, LSTM, and Prophet algorithms. The maximum error achieved by the surrogate model is 3.19°C compared to the validated temperature. The study demonstrates that surrogate models offer a time-efficient alternative for simulating complex physical phenomena, achieving a balance between accuracy and efficiency. This approach is particularly beneficial in scenarios where full-scale 2D and 3D simulations are excessively time-consuming, providing quicker results and significantly reducing computational resources.
Study on the Weights of Influencing Factors Contribution to Background Count Rate in Liquid Scintillation Measurement
Authors: He Shuijun , Tang Lijuan , Liang Manchun , Li Ke , Wang Jihong , Du Xiaochuang 
Published: November 2024

Abstract
The background count rate is a key factor to MDA and accuracy of measuring gross alpha and gross beta in water sample using liquid scintillation analysis. In this paper, the weights of the measurement environment, the material of scintillation vial, the volume of scintillation cocktail, type and volume of sample solvent to the background count rate were studied in Tsinghua University Laboratory and China Jinping Underground Laboratory using liquid scintillation counter. The weight of the external measurement environment is more than 60% to the background count rate in Tsinghua University Laboratory. The wights of the volume of the liquid scintillation cocktail, pure water and 0.05M nitric acid are linearly positive correlation to the background count rate. Pure water and 0.05M nitric acid were used as sample solvent, the weights of the volume of the two kinds of sample solution to the background count rate are comparable. At the same volume, the weights of pure water and 0.05 M nitric acid to the background count rate are lower compared to weight of the liquid scintillation cocktail to the background count rate. Compared with low-potassium glass scintillation vial, plastic scintillation vial has a lower weight to the background count rate. The results can provide data support to reduction of background count rate of liquid scintillation measuring gross alpha and gross beta in water samples, optimization of measuring condition and promotion of detection capability.
Research on Nuclear Power Plant Equipment Layout Method Based on Genetic Algorithm
Authors: Su Jincheng , Jia Xiaopan , Zhao Xiaoshan , Hao Dong 
Published: November 2024

Abstract
Based on the process of nuclear power plant equipment layout, this paper proposes a design method for automatic layout of nuclear power plant equipment in the form of solving a two-dimensional box packing problem based on genetic algorithm. By considering factors and requirements such as the overall dimensions of the nuclear power plant equipment, the process system to which it belongs, the maintenance space of the equipment, radiation levels, etc., a numerical model of the equipment process room is established. Taking factors such as the transportation path of the nuclear power plant equipment layout plan as the main indicators, establish an evaluation system for the nuclear power plant equipment layout plan. Based on the solution to the two-dimensional box packing problem, the nuclear power plant equipment layout process is simplified and a mathematical model for nuclear power plant equipment layout optimization is constructed. The advantage of genetic algorithm in solving optimization problems is used to realize the automatic layout of nuclear power plant equipment and obtain the optimal layout plan. Programming simulation is carried out according to the above method and flow, and under the set equipment layout constraints and genetic algorithm parameters, the automatic layout scheme design of nuclear power plant equipment is carried out and the equipment layout scheme is provided. The relative position of the equipment and the direction and location of the main channels reflected in the plan can provide a design basis for subsequent plan design.
Design Management Platform Maturity Research and Improvement Application
Authors: Shi Haili , Hu Yuefei , Zhao Wenzhao 
Published: November 2024

Abstract
Since its implementation in 2013, the Teamcenter design information management system has served as a comprehensive design management platform for the Nuclear Engineering Research and Design Institute. Over the years, this platform has significantly enhanced the informatization and digitalization of China’s third-generation nuclear power design, equipment manufacturing, and engineering construction. It ensures the advancement, safety, and economic efficiency of the Hualong One project. Nuclear power design and information technology are deeply integrated, allowing various departments to accumulate extensive experience in adapting the platform to suit the business processes, rules, and data of engineering management. However, there is a recognized need for improvements in design management and platform optimization. This paper proposes the establishment of a “design management platform maturity evaluation system” to assess the platform’s effectiveness and visualize evaluation data. Through the definition of evaluation indicators and the establishment of an Analytic Hierarchy Process (AHP) table, this study develops a maturity evaluation system for the design management platform. It analyzes key platform elements via expert scoring to perform the maturity assessment. Subsequent improvements are applied in the CIMS3.0 system, enhancing the safety, efficiency, and convenience of data querying and statistical analysis. This includes features such as quick access to statistical reports and comprehensive data queries, thereby improving the quality and efficiency of design management.
Irradiation-Thermal-Mechanical Coupling Behaviors of a Typical TRISO/SiC Fuel Element Based on Actual Distribution
Authors: Liangjie Zhang , Ping Chen , Shichao Liu , Haoyu Wang , Chong Wei , Xiaoqiang Pan 
Published: November 2024

Abstract
TRISO (Tristructural isotropic)/SiC Fuel Elements is developed and considered a kind of high-temperature gas cooled reactor (HTGR) fuel, it helps to improve the safety, competitiveness, and economics of commercial nuclear power. This study focuses on a typical fuel element, which is manufactured by ceramic SiC and contains more than 10000 surrogate TRISO particles. Then the actual position information of particles is applied in finite element analysis. The finite element model, considering comprehensive coupling behavior (including irradiation-induced dimensional change, irradiation-induced creep, thermal expansion, fission gas pressure, heat generation and conduction, elasticity and plasticity, etc.) and particle to particle/particle to matrix interaction, were developed based on the COMSOL platform and accurately verified. The verification process mainly based on the IAEA benchmark of single TRISO particle, a three-layer model and the fission gas release result using BISON. Based on the single particle model, a more complicated FEM is considered. It provides detailed modeling and analysis of the performance of each layer, which may occupy a huge amount of computing resources. Therefore, representative volume element (RVE) is selected, modeling the geometric structure with symmetry. The result indicated there is a significant high stress in the densely packed areas of fuel element, which increased the failure possibility of particles. By analyzing the stress field, temperature field and deformation of the model, it is indicated that in order to reduce the failure possibility of TRISO particle, the particle clusters should be avoided during fabrication process.
Study on Sodium Extraction and Transport at Monju
Authors: Isobe Yuta , Tanigaki Takanori , Tone Kohei , Joboji Yuya , Matsui Kazuaki , Obata Ikuhito 
Published: November 2024

Abstract
The prototype fast breeder reactor “Monju” is a loop-type, sodium-cooled nuclear reactor in Japan. Its operator, Japan Atomic Energy Agency (JAEA) started decommissioning of Monju in 2018, and the first phase of the decommissioning was completed in FY2022 when all the fuel assemblies were removed from the reactor core and Ex-vessel storage tank (EVST) and transferred to the spent fuel pool (SFP). In FY2023, we started the second phase, and are currently working to remove neutron shields and other core elements from the reactor core and EVST, to put them in SFP. We are also dismantling water and steam system such as turbine generator, condensers and feed water heaters, etc.
MONJU has a primary sodium cooling system for the reactor core and its fuel, and a secondary sodium system that transferred heat. Since the sodium in the primary system is radioactive while the sodium in the secondary system is not, we had developed plans to drain/extract non-radioactive sodium first. According to these plans, we defined the sodium extraction area first, to set up International Organization for Standardization (ISO) tanks for non-radioactive sodium and developed a route for extracting sodium from the existing tanks to the ISO tanks. New piping for these operations will be installed, so that the sodium can be extracted using existing gas systems and electromagnetic pumps. We are ready to take appropriate measures to prevent sodium leakage and continue to develop methods, ensuring safety, to achieve efficient extraction and transport of the sodium.