bluuebunny/crossref_metadata_2025_split huggingface dataset:
Internal Nitridation of Ni-Based Superalloy IN617 During Aging at 950 °C
Authors: Wang Yue , Zhao Hongsheng , Wang Haitao , Shi Li , Li Kejian 
Published: November 2024

Abstract
Aging experiments were performed on Ni-based superalloy IN617 at 950 °C in air (containing 78.8 wt.% N2) for three different durations (50 h, 500 h, and 3000 h). Internal nitridation was found. The layered distribution structures of oxides, nitrides, and carbides were investigated. The formation of oxides and nitrides and the evolution of carbides during aging were analyzed. Results showed that oxidation only occurred in the nearest-surface vicinity, forming an external Cr2O3 layer and an internal Al2O3 layer as well as a Cr-depleted zone. With increasing aging duration, the oxide layer became continuous, and the thickness of the Cr-depleted zone was increased. Nitridation occurred in the sub-surface vicinity below the oxide zones. As aging duration was increased, there were increases in both the thickness of nitride zone and the nitride type. Specifically, only TiN was formed after aging for 50 h. When the aging duration was increased to 500 h, TiN and AlN both precipitated. TiN, AlN, and Cr3.25Ni1.75N all precipitated for the aging duration of 3000 h. At the zones far away from the external surface, the carbides including Cr-rich M23C6 and Mo-rich M6C were coarsened and transformed to discontinuous large blocks from the continuous thin film of the as-received condition under the effect of high temperature.
Research on Fuel Pebble Modeling Methods Comparison and Eigenvalue Analysis of the Running-in Phase of HTR-PM
Authors: Yang Qianye , Gui Nan , Tu Jiyuan , Jiang Shengyao 
Published: November 2024

Abstract
The High-Temperature Gas-Cooled Pebble bed Reactor (HTR) utilizes fuel pebbles that consist of an outer graphite shell and a fuel region with thousands of TRi-structural ISOtropic (TRISO) particles dispersing in a graphite matrix, causing the double heterogeneity effect, and presenting computational challenges for neutronics simulations. The Monte Carlo (MC) method is considered a powerful tool for neutronics analysis of complicated reactor design, but it is computationally expensive and inefficient for large-scale systems. This study calculates and compares different fuel pebble modeling methods to balance accuracy and efficiency. Models ranging from high-fidelity to various homogenized approaches are assessed. Results indicate the Reactivity-equivalent Physical Transformation (RPT) model optimally captures heterogeneous effects while reducing the computational cost. Based on RPT, eigenvalue analysis is performed to simulate the running-in phase of the HTR-PM. A linear correlation is observed between the descending surface height and keff, which is consistent with the phenomenon in reactors of smaller scales like HTR-10. This comparative modeling and analysis provide insights into fuel pebble homogenization and support the feasibility of applying proper fuel pebble modeling to full-core HTR design and simulation.
Core Design and Neutronic Analysis of the European Sodium Fast Reactor With Metallic Fuel
Authors: Jiménez-Carrascosa Antonio , Mikityuk Konstantin , Stauff Nicolas , Karahan Aydin , Fridman Emil , Ponomarev Alexander 
Published: November 2024

Abstract
The current ESFR (European Sodium Fast Reactor) design was proposed and in-depth evaluated in the frame of the past ESFR-SMART project. As a follow-up project, the ESFR-SIMPLE has been launched with the aim of challenging the current commercial-size ESFR design in terms of safety features and economic performance. Among the new safety measures to be developed and assessed in ESFR-SIMPLE, the current oxide fuel ESFR design will be challenged by a modified version of the core with metallic fuel. This intends to conclude on what types of benefits can be obtained with high-density fuel, under similar safety and design constraints. In this paper, the designing approach for enabling the use of metallic fuel in the current ESFR core is described and a preliminary neutronic evaluation is carried out. The optimal configuration is established through the optimization of key neutronic parameters aiming at the potential reduction of the plutonium inventory. The resulting core configuration serves as a basis for further safety assessment analyses, which will provide insight into the advantages and drawbacks of the two types of fuels.
Preliminary Investigation of THFR Passive Residual Heat Removal System
Authors: Huang Yuan , Xu Wei , Xie Heng , Shi Lei 
Published: November 2024

Abstract
The RELAP5 code was employed to study the natural circulation characteristics and core thermal-hydraulic parameters of THFR (Tsinghua High Flux Reactor) under the condition that only the passive residual heat removal system was used when the emergency cooling pump powered by the battery bank shut down due to power exhaustion. To achieve a higher neutron flux, high flux reactor possesses several distinctive properties, including high power density, narrow coolant channels, and low coolant pressure., so the loss of heat sinks under accident conditions can be catastrophic, Therefore, it is necessary to establish a passive residual heat removal system. This study compared the two-loop and three-loop schemes, The results showed that both schemes can reliably establish natural circulation and efficiently remove core residual heat. Throughout the process, the maximum temperatures of the coolant, cladding surface, and fuel meat were below the design limits. Because the maximum temperature of the cladding surface was lower than the saturation temperature, the DNB phenomenon did not appear in the core. The three-loop scheme, with lower total resistance and higher coolant flow rate, is more favorable for improving the thermal safety of the reactor. It also has a lower overall heat transfer area, which helps reduce the cost of heat exchangers.
Development of the AI-Assisted Thermal Hydraulic Analysis Method for Condensing Bubbles in Vertical Subcooled Flow Boiling
Authors: Zhou Wen , Miwa Shuichiro , Okamoto Koji , Okawa Tomio , Tsujimura Ryoma , Nguyen Thanh-Binh 
Published: November 2024

Abstract
Subcooled flow boiling plays a pivotal role in various industrial applications, including nuclear reactors and thermal management systems. However, the rapid and complex changes of condensing bubbles experience from their inception to collapse in subcooled flow boiling present substantial obstacles for conventional bubble detection methods, especially when it comes to condensation bubbles. In light of this, a state-of-the-art AI method is developed and validated for the efficient detection and tracking of condensation bubbles in subcooled flow boiling, thereby enabling the effective execution of thermal hydraulic analyses. This study initially employs computer vision technology to efficiently construct a bubble dataset. Subsequently, a bubble detection model is trained based on this dataset and the YOLOv8 with attention mechanism. Utilizing the results of bubble detection, a multi-object tracking algorithm is then applied to track all bubbles in each frame. The developed AI-based method exhibits a strong ability to detect 95% of condensation bubbles in subcooled flow boiling, additionally, it streamlines the extraction of critical thermal hydraulic parameters, including void fraction, bubble lifetime, and nucleation site density. The model’s accuracy and consistency is demonstrated compared to empirical correlations, affirming its reliability in analyzing the intricate dynamics of subcooled flow boiling.
A Novel Optical Chamber Design and Shielding Optimization for Application in Liquid Scintillation Counting
Authors: Wang Husheng , Zuo Liangzhou , Yan Yangyang , Jiang Hao , Wang Jie 
Published: November 2024

Abstract
Liquid Scintillation Counting (LSC) is a technique widely employed for measuring the activity of radioactive isotopes, finding applications in fields such as radioactive isotope measurements, nuclear material analysis, environmental monitoring, and nuclear medicine. This study is based on the Triple-to-Double Coincidence Ratio (TDCR) method within liquid scintillation counting, proposing a novel TDCR-LSC detector structure with a TriL-shaped optical chamber. Compared to traditional TDCR structures, the TriL-shaped optical chamber integrates the U-shaped design from the Pacific Northwest National Laboratory (PNNL) and the conventional TDCR structure. This is achieved by cutting three L-shaped hollow light guide tubes in the shielding, directing the light from the liquid scintillation vial at a 90° angle into three vertically positioned photomultiplier tubes (PMT). The design avoids the need for heavy external shielding around the PMT, and the L-shaped light guide prevents external radiation from directly entering the sample, thereby enhancing the shielding performance. The TriL-shaped design exhibits significantly improved shielding efficiency compared to the traditional TDCR structure. Validation through Geant4 simulations demonstrates that the TriL-shaped design outperforms the traditional TDCR structure in terms of shielding efficiency while maintaining efficient light collection. This research introduces a novel design approach to enhance the performance of liquid scintillation counting detectors.
Modeling Co-Current and Counter-Current Flow: A Performance Evaluation of the TRACE Condensation Model with Non-Condensable and Light Gases
Authors: Albdour Samah A. , Addad Yacine , Afgan Imran 
Published: November 2024

Abstract
Efficient steam condensation is crucial for ensuring the secure functioning of nuclear power plants (NPPs) by mitigating the potential dangers associated with excessive pressure and overheating. Nonetheless, the existence of non-condensable gases (NCGs) can obstruct this essential process, forming thermal resistance layers, impeding steam diffusion, and hindering condensation on the system’s surfaces. Consequently, the objective of this study is to enhance our comprehension of steam condensation by assessing its effectiveness in the presence of non-condensable gases. The current work utilizes a condensation setup in KAIST’s passive containment cooling system (PCCS) facility, chosen for its validation accessibility. Unlike prior validations focused on pure steam and air using RELAP5/MOD3.2, this study employs TRACE validation to explore the effects of non-condensable gases, including air, nitrogen, hydrogen, and helium. The research generates diverse scenarios, correlating them with TRACE parameters to understand the impact of gases. Using the user-friendly Symbolic Nuclear Analysis Package (SNAP) graphical interface, a geometric model is built and verified in the TRACE code, aligning well with experimental data. Subsequently, 2,100 TRACE cases are generated to evaluate the condensation heat transfer coefficient (HTC). Pearson correlation coefficients from the data highlight that mass fraction has the most significant adverse impact on the HTC, followed closely by the gas type based on molecular weight. Given these notable negative effects, the initial focus is on how mass fraction and molecular weight influence the liquid generation rate in co-current flow condensation. Results demonstrate that a higher mass fraction and molecular weight reduce the liquid generation rate compared to pure steam. Examining their impact on drain mass flow rate in steam/NCG mixture variations reveals their role in triggering counter-current flow limitations (CCFL). Findings suggest that reaching the steam/NCG mixture flow rate threshold stops condensation, initiating CCFL. Moreover, increased mass fraction and a shift to gases with higher molecular weights lead to an earlier onset of CCFL.
C-14 in Geological Repositories: Source Terms and Transport Models
Authors: Qu Zhengzhe , Guo Jingni , Zhang Qi , Wang Yu , Wu Mengjie , Xie Feng , Wei Liqiang , Cao Jianzhu 
Published: November 2024

Abstract
Spent fuel management is one of the key issues for the sustainable development of nuclear energy. The deep geological disposal of radioactive waste is a vital aspect of spent fuel management, aiming to isolate radioactive sources from humans and circulating groundwater as much as possible. Buffer materials, as a crucial part of the engineering barrier system, play an important role in preventing the infiltration of groundwater and the outward migration of radioactive nuclides. However, due to a variety of unforeseen circumstances, there remains a risk of leakage in geological repositories, potentially leading to the radioactive sources penetrating the engineering barrier system and causing environmental damage. Carbon-14 (C-14) is one of the significant radioactive elements in China’s geological repositories. In the event of repository damage, C-14 could enter the biosphere and infiltrate the human body through the food chain. It could pose internal radiation hazards to human tissues and organs, which becomes a major source of collective radiation dose for humans.
In recent years, there has been significant progress in the study of buffer materials for high-level radioactive waste (HLW) geological repositories in China, including the development of GMZ01 type bentonite as a buffer material. Extensive research has been conducted on the basic physicochemical properties and nuclide adsorption characteristics of these materials. However, studies specifically related to C-14 in these repositories are less common. The methods for assessing the inventory of C-14 in the repositories need to be refined. Although some computational models for simulating the migration process of C-14 from the repositories to the environment have been proposed, there is a lack of comprehensive summaries. Additionally, the overall research framework for the migration process of C-14 in buffer materials is not clear.
Therefore, the study of the C-14 source term in geological repositories and its migration and diffusion behavior in bentonite, such as GMZ01 type, is a crucial part of the radiation safety assessment in geological repositories in China. This paper focuses on C-14, which is a typical radionuclide in radioactive source terms in geological repositories, investigates the research progress on the C-14 source term in geological repositories both domestically and internationally, organizes the methods for assessing the inventory of C-14 in repositories and compiles computational models for the migration and diffusion behavior of C-14 in key buffer materials. This paper aims to provide theoretical support for the analysis of the C-14 source term in high-level radioactive waste geological repositories and the study of C-14 migration processes in buffer materials, which could provide an important research basis for the radiation safety evaluation of high-level waste geological repositories in China.
Fault Detection and Location Method for Nuclear Power Plant Main Systems Based on Low-Dimensional Manifolds
Authors: Ai Xin , Liu Yongkuo , Shan Longfei , Gao Jiarong 
Published: November 2024

Abstract
The nuclear power plants have characteristics such as variable operating conditions, many large and complex system, high-dimensional fault data, and complex nonlinearity. Classical fault diagnosis methods such as BP neural networks and support vector machines lack excellent feature extraction capabilities and have limited data learning abilities, making them unsuitable for multi-fault diagnosis tasks in complex, nonlinear nuclear power plants with multiple parameters. To address these issues, this study proposes a multi-fault diagnosis model for the main system of nuclear power plants based on low-dimensional manifolds. Following the concept of distributed diagnosis, the study divides the large and complex main system of the nuclear power plant into several subsystems, transforming the complex multi-system diagnosis task into simpler fault detection and localization tasks for each subsystem. Subsequently, within each segmented subsystem, considering the strong nonlinear coupling relationships among fault features and the difficulty in feature extraction, the study employs a stacked autoencoder method to reduce the dimensionality of fault data and uses the random forest method for classification to achieve fault detection and localization in each subsystem. Through simulation data testing, the fault diagnosis method developed in this study proves to be feasible, providing a reference for addressing the large-scale fault diagnosis challenges in complex nuclear power plants.
Progress on the Development of DAYU3D Code for HTR Thermal-Hydraulic Design and System Analysis
Authors: Zhang Haojie , She Ding , Shi Lei 
Published: November 2024

Abstract
System analysis code is an important tool for thermal-hydraulic design and accident analysis of nuclear reactors. At present, the commonly used system analysis codes for pebble-bed high-temperature gas-cooled reactor (HTR), such as THERMIX and TINTE, adopt two-dimensional (2D) models in the reactor core. A new three-dimensional (3D) thermal-hydraulic design and system analysis code named DAYU3D is being developed by the Institute of Nuclear Energy and New Energy Technology (INET) at Tsinghua University. DAYU3D is able to calculate 3D solid temperature, one-dimensional (1D) multi-batch fuel element temperature, 3D gas flow, and 3D gas temperature. Porous media model is employed to perform the 3D temperature and flow field calculations for the entire reactor. In the pebble bed core, DAYU3D employs 1D thermal conduction model to accurately obtain the temperature distribution inside the fuel element. By comparing calculation results with SANA experimental data, the 3D solid temperature module of DAYU3D in steady state is validated. Also, the three-dimensional and multi-batch calculation capabilities of DAYU3D are demonstrated through the calculation of a sample steady state case for pebble-bed reactor. It is expected that DAYU3D will provide an excellent platform for the research of complex 3D problems and facilitate the precise design of HTR.
Experimental Investigation on the Survival of Cooling Tube in Molten Pool
Authors: Pei Jie , Quan Fengyang , Li Wei , Yuan Yidan 
Published: November 2024

Abstract
An experimental facility was established to verify the structural integrity of the cooling tube under an intense thermal shock formed by the molten corium. The vertical cooling tube specimen made by stainless steel has a diameter of 80mm and a thickness of 10mm, and it is positioned at the center of the test section. The bottom of the tube was connected with a water tank, then the water level and temperature in the tube could be adjusted by changing the relative height of the tank, and the top of the tube was opened to discharge the vapor to the air. The C/K-type thermocouples located at the outer-surface of tube were used to record time that melt interacted with tube, also used to estimate the deep of molten pool. In this study, the simulant material ZrO2 was melted by using the cold crucible technique, then the liquid melt about 70kg was discharged into the test section. The experimental results implies that as soon as the melt contacted with the cooling tube, the ZrO2-crust with the low thermal conductivity was formed around the tube, and it will be significantly decreased the temperature on the inner and outer wall of cooling tube. So, the local maximum temperature in outer wall was near the melting point of stainless steel, the structural integrity of cooling tube was still kept.
Analysis of Mobile Equipment Configuration for Severe Accidents in Nuclear Power Plants Based on Multi-Unit PSA
Authors: Liu Dalin , Wang Jiangguo , Yan Zhen , Wang Shuting , Zhao Ling 
Published: November 2024

Abstract
Multi-unit safety analysis is of great significance for the safety improvement of most nuclear power plants. Meanwhile, the configuration of mobile equipment for severe accidents is diverse in different nuclear power plants, a mature evaluation method is urgently needed in the industry. The configuration evaluation method proposed in this paper can effectively solve the pain point of insufficient theoretical analysis basis and the lack of systematic and effective method when configuring mobile equipment for severe accidents in nuclear power plants. It can help plant owners to make configuration plans according to their own risk conditions avoiding over- or under-allocation, and improve overall safety while reducing procurement costs. The method is basically applicable to all operating reactor types and all nuclear power sites, which has good universality. It is also targeted and specialized for the specific analysis of each nuclear power plants. In addition, a modeling methodology is developed in this paper to build a MUPSA model for level-1 internal events based on existing SUPSA models, which is then applied to the analysis of mobile equipment configuration for severe accidents at a pilot multi-unit nuclear power site.
Numerical Simulation of Subchannel Flow Boiling Using Five-Component Wall Boiling Model and iMUSIG Model
Authors: Luo Hanwen , Wang Hongbin , Xiong Jinbiao 
Published: November 2024

Abstract
Boiling heat transfer of subchannel has significant influence on nuclear reactor thermal-hydraulic performance. Employing the Euler-Euler two-fluid approach in computational multi-fluid dynamic (CMFD) simulations, this study conducts a high-fidelity multi-fluid simulation on subchannels. Utilizing the OECD/NRC NUPEC PWR Subchannel and Bundle Tests (PSBT) benchmark, four subchannels are numerically simulated: a typical central subchannel, a central subchannel with a thimble, a side subchannel, and a corner subchannel. The simulation model integrates a five-component wall boiling model and an inhomogeneous multiple size group (iMUSIG) model in OpenFOAM v6, capturing phenomena like bubble sliding, size variability, coalescence, and breakup. Turbulence is modeled using the standard k-ϵ and Sato models, the latter incorporating an additional turbulent viscosity term for bubble-induced turbulence. The study compares cross-sectional averaged void fractions at measurement planes and analyzes key bubble dynamics parameters, including bubble departure and lift-off diameters, wall heat flux partitioning, and bubble size distribution. The results have shown the promising accuracy of the five-component wall boiling model coupled with the iMUSIG model in predicting the void fraction.
Simulation Study of Transient Characteristics of Liquid Lead-Bismuth Natural Circulation Loop
Authors: Sheng Zhenhua , Yang Zhen , Geng Tengjun , Wang Shengfei , Lv Haicai , Guo Wentao , Liu Fang , Niu Fenglei 
Published: November 2024

Abstract
To study the natural circulation capabilities of Liquid Lead Bismuth (LBE), natural circulation loops have been designed and constructed in several countries for natural circulation characterization of LBE. These loops aim to characterize LBE’ natural circulation properties. Seeking a more profound comprehension of LBE behavior during natural circulation, this research leverages a Computational Fluid Dynamics (CFD) simulation program. This program theoretically models the circulation loop, while a one-dimensional code is employed to verify the model’s accuracy. The findings reveal that the discrepancies between the two models are minimal, thereby affirming their trustworthiness. Utilizing the validated model, a deepening of the influence of loop geometric parameter, heating power, and the configuration of the pooled solid-state oxygen control device on the natural circulation flow rate within the lead-bismuth loop. Furthermore, the CFD model is instrumental in examining the thermal and hydraulic attributes of the loop under distinct operational scenarios, namely transient startup and power jump conditions. In essence, this paper integrates two robust models to explore the thermal and hydraulic nuances of the liquid lead bismuth recirculation loop. Through this exploration, the optimal geometric parameters are established and an assessment of the viability of the pool-type solid-state oxygen control device is also performed.