bluuebunny/crossref_metadata_2025_split huggingface dataset:
Criticality and Burnup Analysis of Accident-Tolerant Fuels in High-Temperature Test Reactor
Authors: Chen Yan-Xin , Wu Shin-Rong , Chao Der-Sheng , Hong Jhao-Yang , Liang Jenq-Horng 
Published: November 2024

Abstract
This study aims to investigate the feasibility of implementing accident-tolerant fuels (ATFs) in high-temperature gas-cooled reactors (HTGRs). Endeavors to reinforce risk management in nuclear safety have led to an increased emphasis on investigating ATFs. These specialized fuels are engineered to endure higher operational temperatures than conventional fuel rods, thereby offering superior thermal stability during accidents. In this study, all the computation was performed by using the MCNP6.2 computer code in conjunction with ENDF/B-VII continuous energy library. Five distinct fuel configurations were investigated: the original UO2 tri-isotropic (TRISO) particle design, UC and UN fuels with TRISO particle design; UC fuel with titanium nitride (TiN) coating, and UN fuel with zirconium carbide (ZrC) coating. The effective multiplication factor (keff) and the burnup performance during 660 equivalent full power days (EFPD) was computed to prove the advantage of utilizing higher uranium density ATFs, with the potential to enhance fuel cycle economy. Additionally, the neutron flux, moderator-to-fuel ratio and temperature coefficient was presented and analyzed in this paper.
Study on Development Pathways of Nuclear Energy for Energy System in China Using Message Model
Authors: Mei Luhan , Chen Xiaotong , Zhang Huang 
Published: November 2024

Abstract
Global climate change is a challenge to human society, and major countries around the world are taking a carbon-neutral pathway. China introduced the carbon peaking and carbon neutrality goals in 2020, highlighting the criticality of reducing carbon emissions. Nuclear power emerges as an essential option for a stable electricity supply with extremely low carbon emissions. This work conducts research on the development pathway of nuclear energy based on the MESSAGE model. Taking Jiangsu province in China as an example to elaborate on the scale, and path of China’s nuclear energy development from the present to 2060.
The model simplifies the representation of coal-fired power, nuclear power, and renewable sources in Jiangsu Province’s electricity system, considering constraints such as demand, generation capacity, regional characteristics for nuclear plant site selection, and carbon neutrality goals. The objective is to minimize costs while meeting electricity demand and other constraints to provide a comprehensive depiction of the nuclear energy landscape. The computational solution yields future activity levels, capacities, and additional capacities for different technologies.
The research results indicate that the future development trajectory of nuclear energy will vary significantly under different scenarios. Against the backdrop of dual carbon goals, nuclear energy is poised to become a crucial component of the future energy system, holding a stable share of the electricity system. Coal-fired power will gradually phase out its dominant position, giving way to new energy plants as a primary component of the future energy system.
This work proposes a quantitative tool that uses an optimization model to analyze the dynamic competition of nuclear energy technology. Through the analysis of model results, scientific planning for future nuclear energy development can be conducted, providing theoretical support for policy formulation and the planning of nuclear energy development paths.
Assessment of the Force-Balance Model in Predicting the Bubble Departure Diameter in Inclined Channels
Authors: He Wen , Han Jinyu , Zhao Chenru , Bo Hanliang 
Published: November 2024

Abstract
The flow boiling is widely encountered in various fields, ranging from the chemical and process industries to boiling water reactors and steam power plants etc. In order to gain a better understanding of this boiling process, one of the bubble dynamic parameters, the bubble departure diameter is studied in this paper. Firstly, the forces acting on a nucleation bubble were analyzed based on the force balance model, including the surface tension force, unsteady drag force, shear lift force, buoyancy force, hydrodynamic pressure force, quasi-steady drag force and contact pressure force. Then, a large number of experimental data for bubble departure diameter in flow boiling conditions under a wide range of orientation angle, mass flux, heat flux, and pressure were collected to quantify the predicting accuracy of existing force balance models. Considering that the influence of forces acting on a vapor bubbles would change with the channel inclination, these experimental data were divided into three parts based on the channel inclination. The verification results are of great importance for accurately determining the bubble departure diameter as well as analyzing the heat transfer process under various channel inclination angels.
Study on Friction and Wear Performance of Flow-Blocking Packing for High-Temperature Gas-Cooled Reactor
Authors: Wang Dexin , Min Qi , Ma Yuanyuan , Shi Li , Li Xiaowei , Zhang Zhengming , Sun Libin 
Published: November 2024

Abstract
To mitigate the bypass helium flow within the core structure of a high-temperature gas-cooled reactor and enhance the efficiency and power output, this study explores the use of braided packing as flow-blocking components in the core support structure. Structural design and optimization were carried out, considering various technical factors associated with the application of braided packing as fillers in the high-temperature gas-cooled reactor core support structure.
Four distinct models of braided packing were compared for their friction and wear performance under diverse environmental conditions, including high temperatures and a helium atmosphere. Possible friction and wear mechanisms, along with influencing factors, were investigated. Through the evaluation of data such as loss in mass after wear, coefficient of friction, and post-wear surface morphology, preliminary conclusions suggest that braided packing may meet the operational requirements for a 60-year lifespan of a high-temperature gas-cooled reactor.
The test results indicate that the selected braided packing exhibits favorable friction and wear performance in the high-temperature gas-cooled reactor core environment. The loss in mass after wear, under frictional distances equivalent to a 60-year lifespan of a high-temperature gas-cooled reactor, is deemed acceptable. Furthermore, friction and wear performance testing was conducted on braided packing subjected to 450°C thermal aging for 2000 hours. The experimental outcomes reveal that the braided packing, after undergoing 450°C thermal aging for 2000 hours, still demonstrates satisfactory performance. Based on the aforementioned research findings, it can be concluded that disk materials, when employed as flow-restricting components, do not experience functional failure due to friction and wear mechanisms in a high-temperature gas-cooled reactor.
Research and Application of Projection-Based Augmented Reality (PAR) System in Fusion Tokamak Assembly
Authors: Dong Cuicai , Xu Hanjie , Han Wei 
Published: November 2024

Abstract
Projection-based Augmented Reality (PAR) systems have been applied in various industries. Detailed drawings and markings can be easily projected with laser onto the workpiece surface, providing clear instructions to the assembly operators, which brings benefits in efficiency and quality improvement. In nuclear fusion Tokamak assembly work, millions of attachments for diagnostics, cables, and monitoring systems need to be installed on the surface of the Vacuum Vessel (VV) requiring top-level assembly accuracy. Thus, difficulties have been forecasted in marking and component alignment in traditional assembly methods. This paper focuses on a case study of using the PAR system in VV assembly work which demonstrates the improvement in installation efficiency and quality. It showed that the total alignment tolerance can be controlled within 0.2mm, satisfying the assembly tolerance requirement of 2 mm. Installation performance, such as non-conformities and working man-hours, has been statistically improved compared with the traditional installation method. Over 80% of templating man-hours are eliminated for the contractor, with millions of costs saved. It is concluded that with the help of the PAR system, the fusion Tokamak machine assembly can be more efficiently conducted.
A Policy Driven Vegetative Foodchain Model: Soil-Plant Transfer in Chinese Organic Soil
Authors: Oforiwaa Priscilla , Zhang Xiaole , Su Guofeng 
Published: November 2024

Abstract
Radionuclides can be released into the terrestrial environment from nuclear facilities through particulate matter or gases. The deposition of radionuclides in vegetation and soil is the starting point for their transfer into the terrestrial environment and food chain. Two main depositional processes remove pollutants from the atmosphere. Interception is defined as the portion of radionuclides deposited by wet and dry deposition that is initially retained by vegetation. While the retained radionuclides are subsequently weathered into the soil, the initially intercepted fraction is a very significant quantity in any radioecological model. Once deposited on vegetation, radionuclides are lost from plants by wind and rain through leaching or abrasion of the cuticle. The increase in biomass during growth does not result in a loss of radioactive material; however, due to effective dilution, it can lead to a decrease in the radioactive concentration. Transfer describes the systemic transport of radionuclides following foliar uptake by plants. However, for estimating the concentration of radionuclides in food and assessing doses to humans, the systemic transport of radionuclides is a critical issue. This is especially important for plants where only certain parts are used as food or feed, such as cereals and potatoes. Food chain modeling development is a vital part of emergency planning and response in the fields of nuclear industries, chemical industries, petroleum, and fire.
This study reports on the evolution of a novel vegetative model, the estimation of radionuclide concentrations in organic soil, and radionuclide migration through the food chain using compartment model calculations, to review China’s policy on nuclear safety and develop a model to obtain data that will contribute to safety assessment. To validate the accuracy of the model, the simulated concentrations of radionuclides were compared with the results of an existing model ECOSYS-87. The mathematical model for the simulation was developed using Matlab a mathematical programming tool developed by Math Works. The radionuclides considered for the study are Cs-134, 137, Iridium-131, and Strontium 89, 90. The results indicate a higher concentration of tubers and herbs with the least concentration in woody plants such as fruits.
A Study of Safety Importance Screening Criteria in the NPP Risk-Informed Safety Classification Process
Authors: Zhou Jinghua , Deng Wei , Ma Chao , Guo Yiwen 
Published: November 2024

Abstract
Long-term practical operational experience and analysis of nuclear power plants have shown that component classification based on determinism can be optimized. Some safety-related components are not important from a risk perspective, while some non-safety-related components may have a large impact on the safety of the plant. With the development of probabilistic safety analysis (PSA) and the application of risk-informed technology, a risk-informed classification method combined deterministic and probabilistic evaluation has been developed. Risk-informed safety classifications (RISC) process can help nuclear power plants to focus their resources on high safety-significance component, thereby improving plant safety. By relaxing the regulatory requirements for component of low safety-significance, the operating cost can be reduced to a certain extent and the economic benefits of the power plant can be enhanced. The first step of RISC process is risk characterization of SSCs, using the sum of Fussell-Vesely (F-V) importance and Risk Achievement Worth (RAW) values as measurement parameters. F-V importance and RAW examine the impact of events on nuclear power plant risk from different perspectives and are commonly used risk importance parameters in risk quantification work. However, in Chinese engineering practice, Risk Spectrum (RS) software is often used to calculate the risk importance of basic events, and by creating component groups, it can directly give the risk importance of component. In this paper, we will analyze the calculation of the component risk importance provided in the RS software, using the PSA models of a generation II reactors and a generation III reactor as example, compare it with the measurement parameters provided by the NEI 00-04. In addition, by collating the risk importance distribution of nuclear power plant component, this paper makes some suggestions for screening metrics, which provide a reference for the subsequent pilot work of risk-informed component categorization in operating power plants.
Structural Design and Analysis of Plate-Type Fuel Assembly of Research Reactor
Authors: Kim Hyun-Jung , Tahk Young-Wook , Kim Dong-Hyun 
Published: November 2024

Abstract
Plate-type fuel assembly is widely used in research reactor because of its long-term stable behavior of fuel performance and excellent thermal-mechanical behavior. KJRR (Ki-jang research reactor) fuel assembly adopted plate-type fuel which consists of U-Mo dispersed fuel meat and aluminum matrix. In structural design point of views, fuel structural integrity shall be maintained during the normal operation condition, AOOs (Anticipated operational occurrences) and non-operation conditions, and coolable geometry of fuel assembly shall be maintained during accident such as SSE (safe shutdown earthquake). Structural analyses of KJRR fuel assembly are performed considering the various structural loads expected in the reactor environment. To show that structural damage does not occur due to vibration, a flow-induced vibration test is performed in the simulated reactor loop condition. When the flow velocity is raised to sufficiently high value, fuel plates can be buckled due to unstabilizing hydrodynamic force. The measurement test of critical velocity where the buckling instability begin to occur is performed. This paper presents the structural design criteria of plate-type fuel assembly and summarizes the structural evaluation results.
Properties of Irradiated Additively Manufactured 316L Steel for In-Core Applications – Mechanical and Microstructural Analyses
Authors: Jedlan Štěpán , Ševeček Martin , Hodek Josef , Prantl Anotnín , Šoltés Jaroslav , Fedoriková Alica , Sovadina Michael 
Published: November 2024

Abstract
This paper focuses on separate aspects of the utilization of additive manufacturing (AM) for the in-core applications in the nuclear power industry. This study investigates the application of the laser direct energy deposition method (L-DED) to produce in-reactor metallic components within LWR nuclear reactors. 316L stainless steel with heat treatment was chosen as the material of interest for experimental evaluation of the material properties after the neutron irradiation. Samples for the tensile test at 300 °C and microstructural analysis (LOM, EBSD, SEM) were irradiated inside a drop-in capsule in the LVR-15 research reactor at Research Center Rez. Subsequently, the postirradiation examination (PIE) results are compared to those of non-irradiated counterparts to analyze the impact of in-reactor conditions. Furthermore, the non-irradiated samples, both conventional and AM, were tested at both the Research Center Rez and COMTES FHT a.s, to obtain independent results and validate the methodologies.
To design the irradiation campaign, the calculation of residual activity after the irradiation period is performed using a SCALE module ORIGEN to inform future handling procedures and scheduling of post-irradiation examinations. The calculation of expected displacement per atom (dpa) is also included for direct comparison with results available in the literature.
Research and Preliminary Verification of the Resonance Self-Shielding Calculation Method for the High-Fidelity Neutronics Code VITAS
Authors: Yin Han , Zhang Tengfei , Liu Xiaojing 
Published: November 2024

Abstract
VITAS is a high-fidelity neutronics code developed by Shanghai Jiao Tong University based on the heterogenous variational nodal method (HVNM). In previous publications, VITAS has been successfully verified for transport calculations against macroscopic benchmarks. However, advancing the research on the resonance self-shielding calculation method within VITAS remains crucial for efficient one-step neutronics calculations. This paper presents the implementation of the subgroup method in VITAS. Subgroup parameters are obtained through least squares fitting based on the resonance cross-sections table. The narrow resonance approximation elastic scattering model is employed for the scattering source term, and a subgroup transport equation is established. The subgroup transport equation is solved using the HVNM, yielding subgroup fluxes. Effective resonance self-shielding cross-sections are obtained using the subgroup fluxes to weight the subgroup cross-sections. Additionally, the Bondarenko method is applied to address multiple resonance nuclide interference effects. This paper conducts preliminary verification of resonance self-shielding calculation method through a series of benchmark problems. The numerical results suggest that the effective resonance self-shielding cross-sections and eigenvalue obtained from VITAS show good agreements with the reference solutions provided by Monte Carlo code OpenMC. This consistency demonstrates the capability of VITAS in addressing heterogeneous resonance self-shielding problems.
Enhancing Image Quality in Limited-Angle CT Reconstruction for Reactor Pressure Pipelines Detection
Authors: Fu Jintao , Chang Jiahao , Sun Yuewen , Zeng Tianchen , Tang Peng , Xu Shuo , Cong Peng 
Published: November 2024

Abstract
The pressure pipelines of a nuclear reactor play a pivotal role in ensuring the structural safety of the reactor, and the accurate detection of internal and surface defects constitutes a critical step in safeguarding its integrity. Despite the remarkable performance of CT technology in detecting surface and internal defects, the inspection of reactor pressure pipelines faces significant challenges due to the limitations imposed by limited-angle CT scanning. This paper is dedicated to the investigation of limited-angle CT reconstruction algorithms, conducting an in-depth analysis of traditional iterative reconstruction algorithms such as SART and ADMM-TV. The primary objective is to enhance the image quality by addressing the artifacts caused by the limited-angle acquisition in the detection of reactor pressure pipelines. Experimental results from CT detection of reactor pressure pipelines indicate conspicuous artifacts under limited-angle conditions. Although iterative reconstruction algorithms like SART and ADMM-TV contribute to alleviating some of these artifacts, our proposed DRP reconstruction algorithm demonstrates a more pronounced improvement in reconstruction quality. Through the refinement of limited-angle CT reconstruction algorithms, this paper provides a preliminary exploration of non-destructive testing and reconstruction for reactor pressure pipelines, offering valuable insights to drive technological advancements in the field. The study aims to serve as an important reference for promoting the development of techniques in reactor pressure pipelines detection.
The Effect of Liquid and Wall Properties Toward the Increase of Wetting Velocity in the Case of Using Multiple Plates in the Top-Reflood Vertical Surfaces
Authors: Akbari  , Umebayashi Hiroyuki , Okawa Tomio 
Published: November 2024

Abstract
Rewetting is an important mechanism in the quenching process, particularly in the case of nuclear reactor safety. During the Loss-of-coolant accident (LOCA) in the Emergency Core Cooling System (ECCS) of the nuclear reactor. In the location near the wetting front, heat transfer coefficient (HTC) increases, and the quenching rate also occurs rapidly. The mechanism of wetting velocity on the case of 2 (two) plates is assumed to be different compared to 1 (one) plate.
The test section consists of two metal plates which are used as the heat transfer surface. Each of up and down edges of the metal plates (brass or copper) was sandwiched by the brass block columns. The block columns are set to heat the plate until the temperature of 250–300 °C. Meanwhile, distilled water is sprayed through the nozzle onto one or two plates. Furthermore, wetting velocity is calculated using the video captured by infrared camera FLIR™ to observe the liquid motion on the heated wall.
Several parameters are being observed, including the effect of mass flux, initial wall temperature, thermal conductivity, wall thickness, and subcooled degree. The results show the significance of droplets impact for cooling the wall, thus contributes to higher wetting velocity in the case of multiple plates.