bluuebunny/crossref_metadata_2025_split huggingface dataset:
Investigating the Motion Characteristics of Complex-Shaped Foreign Objects on the Secondary Side of a Steam Generator Based on CFD-DEM Method
Authors: Mu Defang , Rong Jie , Wang Mingjun , Tian Wenxi , Qiu Suizheng , Su Guanghui 
Published: November 2024

Abstract
Steam generator (SG) stands as a pivotal element within the primary circuit pressure boundary of nuclear power plants, functioning as a critical safety barrier against potential radioactive leaks. Throughout its operational lifecycle, the introduction of foreign objects (FOs) with diverse shapes into the secondary side of the SG may occur due to various factors, resulting in wear and tear on the pipes. Throughout its operational lifecycle, the introduction of foreign objects with diverse shapes into the secondary side of the SG may occur due to various factors, resulting in wear and tear on the pipes. To safeguard against substantial damage to the steam generator due to foreign objects, it is imperative to evaluate the distribution patterns of these objects and their potential impact on the steam generator. This study employs the Computational Fluid Dynamics-Discrete Element Method (CFD-DEM) coupling approach to develop a three-dimensional model of a steam generators feedwater ring. We have conducted an in -depth investigation into the movement characteristics exhibited by block-shaped foreign objects. By establishing a model for the interaction between foreign objects and the flow field, and considering the contact model of foreign object-wall interactions, we have achieved accurate prediction of the trajectories and distribution patterns of foreign objects within the steam generators feedwater ring.
A Preliminary Thermodynamic Model of Hydrogen Generation Using Solid Oxide Electrolysis Cell (SOEC) Coupled With a High-Temperature Gas-Cooled Reactor
Authors: Gao Wenxiu , Liu Xiongbin , Zhou Zhende , Li Xiaowei 
Published: November 2024

Abstract
To pave the way for industrial-scale hydrogen production using the high-temperature gas-cooled reactor (HTGR), it is necessary to assess the technical feasibility of various hydrogen production processes coupled with a HTGR. In this paper, several promising hydrogen production methods were compared from the perspectives of technological maturity, economic viability, and coupling efficiency with HTGR plants. Among these methods, the solid oxide electrolysis cell (SOEC) hydrogen production technology was considered the most promising approach due to its high efficiency and ease of industrial-scale production. A preliminary thermodynamic model was established for the HTGR-SOEC co-generation plant based on the parameters of a typical HTGR with a full power of 250 MWe and a steam temperature of 540 °C. The normal operating current and voltage of the SOECs coupled with HTGR were chosen as 1 A and 1.287 v, respectively, and SOEC modules were stacked together in the thermodynamic model to achieve the final hydrogen output. Key parameters that influence the hydrogen production rate and efficiency were simulated and discussed in detail using the thermodynamic model. Current research has deepened the study of carbon-free hydrogen production technology using nuclear energy and provided useful insights for future hydrogen-electricity co-generation applications.
Characteristic Study of Wet Steam Condensation With Different Mathematical Models in High-Pressure and Low-Pressure Nozzles
Authors: Jiang Zhuojun , Sun Wan , Tang Simiao , Pan Liangming , Zhang Luteng , Ma Zaiyong 
Published: November 2024

Abstract
Turbines play a critical role as the primary energy conversion devices in the conventional island of nuclear power plants, efficiently transforming the thermal energy of steam into mechanical energy. They hold a significant position in the power generation system. While thermal power turbines often generate wet steam in the final stages, nuclear power turbines, particularly those in pressurized water reactor (PWR) units, exhibit distinct characteristics. The lower steam parameters produced in the steam generators result in the majority of the working fluid entering the turbine as either saturated steam or steam with a certain degree of moisture. Consequently, compared to thermal power turbines, a substantial portion of the entire turbine set in nuclear power turbines operates within the wet steam region, exacerbating the phenomenon of spontaneous condensation of wet steam.
Previous research has predominantly focused on the last stages of both thermal and nuclear power turbines, with limited attention given to the non-equilibrium condensation of wet steam in high-pressure conditions during turbine operation.
Nozzles and blades, due to their structural similarity to the turbine flow path, are commonly employed as representative components for mathematical modeling and characteristic studies of turbines. In this study, the Euler-Euler two-phase flow mathematical model was adopted, and the nucleation and droplet models were selected and modified. The IWSEP nozzle and Gyarmathy nozzle were used to simulate wet steam flow under low and high inlet pressures, respectively. Through numerical simulations, the results were compared with previous simulation outcomes and experimental data. Additionally, by exploring the impact of different nucleation and droplet growth models, the study investigated the influence of mathematical models on wet steam condensation.
The research yielded insights into the evolution patterns of the flow field inside the nozzle under different condensation models. The results indicate that the numerical prediction method established in this study can be applied to the primary droplet prediction in both high and low-pressure nozzles. This is advantageous in reducing heat losses attributed to wet steam condensation, providing a theoretical foundation and basis for the practical operation of turbines.
Assessment of Multi-Unit Emergency Response Capability of Nuclear Power Plants
Authors: Gao Jiaxuan , Liu Ya , Xue Na , Zhang Jiemin , Wang Mengxi 
Published: November 2024

Abstract
After the Fukushima accident, the multi-unit accident has attracted widespread attention in various countries. However, at present, the assessment of emergency response capability generally focuses on a single unit, and with the construction of multiple units at one site, it is necessary to evaluate whether the site has the ability to deal with multi-unit accidents. Therefore, this paper takes a multi-reactor nuclear power plant site in China as an example to evaluate the emergency response capability of the multi-unit accidents from the perspective of the improvement items of the site after Fukushima, the ability to mitigate the consequences of the accident, the management of the multi-unit site, and the response to the multi-unit accident. The ability to mitigate the consequences of the accident will be analyzed from the perspective of the design of HPR1000. In terms of management, each unit at the site is equipped with a basic on-site emergency organization, and regular emergency training and exercise are conducted for staff in different positions. In order to improve and enhance the emergency ability of multi-unit accidents, this paper suggests that the site should clarify the manpower arrangement and specific responsibilities in terms of emergency coordination between units in the event of multi-unit emergency. In terms of multi-unit accidents response, the site is equipped with emergency rescue materials to deal with multi-unit accidents and external support will provide help if necessary. In general, the multi-unit emergency response capability of the site is good, and it can be further improved in the future in clarifying the responsibilities of multi-unit emergency personnel.
Control Scheme of Cogeneration Nuclear Reactor Plant Based on the “Turbine Follows Heating” Operation Mode
Authors: Zhang Ru , Zhang Nan , Yan Mengxiao , Sun Peiwei , Wei Xinyu 
Published: November 2024

Abstract
Cogeneration is one of the ways to reduce carbon emissions and improve the efficiency of nuclear reactor plants. After adding thermal load, the power balance relationship between the supply and demand is changed in cogeneration pressurized water reactors (Co-PWR). So, the steam extraction from the turbine is chosen to analyze the control scheme in Co-PWR. First, the simulation model is established and verified based on the 3KeyMaster simulation platform. Then, the control scheme is designed based on the operation mode of the “turbine follows heating”. Three control systems are considered in the control scheme, which are the nuclear power control system, turbine control system, and thermal energy control system. Finally, the designed control scheme is verified by the simulation of typical scenes. The simulation results show that the designed control scheme makes the Co-PWR operate with stability and coordination. The study can provide a reference for the design of control schemes in Co-NPPs based on the other operation modes.
Research on Operating Reliability of Multi-Modular High Temperature Gas-Cooled Reactor Nuclear Power Plant Based on Equipment State
Authors: Guo Chao , Li Jianghai , Jia Qianqian , Qu Ronghong , Huang Xiaojin 
Published: November 2024

Abstract
Compared with nuclear power plants (NPPs) with single nuclear steam supply system (NSSS) module, Multi-modular High Temperature Gas-Cooled Reactor (MHTGR) has a larger scale of equipment and more complex human operation activities. It is necessary to study the operating reliability of MHTGR from the perspective of NPP equipment to alert the operator the current status of NPP. In this paper, an MHTGR operating reliability model is proposed based on equipment state. Markov process method is used for equipment or system modelling. The system states are simplified to avoid the state explosion in the model. Based on the sharing forms, the operating reliability model of the whole MHTGR is established with system set classification method. Several operating reliability indices are proposed from multiple perspectives, including power generation capacity, system failure severity, and operating reliability level. The abnormality of HTR-PM600 helium circulator system is used as an example to verify the operating reliability modeling method. The experimental results show that the research can reflect the impact of equipment status on plant operation, which helps operators to identify operation risks, improve overall perception awareness, so as to improve the operation performance of MHTGR NPPs.
Numerical Analysis and Model Evaluation on Flow and Heat Transfer Characteristics of LBE in a Pipe Based on DNS Data
Authors: Ai Zhelun , Li Jie , Sun Yumeng , Miao Heng 
Published: November 2024

Abstract
The liquid lead-bismuth eutectic (LBE) considered as an alternative coolant for liquid metal-cooled fast reactors has the characteristics of high density and ultra-low Prandtl number (Pr) which produces significantly distinctive characteristics compared to ordinary coolants. Numerical simulation study on the flow and heat transfer characteristics in a straight pipe of length 12.5D with constant wall temperature and constant heat flux boundary conditions separately. With the condition of Reynolds number (Re) range of 5300∼37700, the comprehensive analysis based on calculated results was performed. The calculated results of axial velocity profiles, near-wall temperature and turbulent stress are compared with the data calculated by Direct Numerical Simulation (DNS) or Large Eddy Simulation (LES). By comparing multiple turbulence Prandtl number (Prt) models, this study aims to provide advice for the selection of turbulence models suitable for LBE. In addition, the study gives the most effective Prt model for specific conditions. It is noted that multiple Prt models demonstrate varying degrees of applicability for different boundary conditions. While the predictions on axial velocity remain consistent with different Prt models under above working conditions, the agreement on wall temperature compared with DNS or LES data varies.
Numerical Study of Film Cooling and Particle Depositions in Turbine Blade of High-Temperature Gas-Cooled Reactor
Authors: Wang Xiaozhong , Sun Qi , Ye Ping , Ma Wenkui , Yang Xiaoyong , Peng Wei 
Published: November 2024

Abstract
Direct-circulation helium gas turbines represent an important development for high-temperature gas-cooled reactors. During the operation of such a reactor, graphite dust is generated continuously in the core, so it is important to investigate the transport characteristics of graphite dust particles in the helium turbine and the mechanism whereby they interact with the cooling characteristics of the turbine blades. Proposed here is a slotted-hole film cooling scheme to effectively cool the helium turbine, and the motion of graphite particles is analyzed. The discrete phase model is used to calculate the particle trajectories, and user-defined functions are used to establish the deposition and rebound models for the particle-wall interactions. The mechanism whereby the slotted holes improve both the cooling performance and the distribution of film cooling effectiveness is studied, and the deposition rate of graphite dust is calculated for different particle sizes. The results indicate that (i) with increasing blowing ratio, the film cooling effectiveness increases and then decreases, (ii) the deposition rate of smaller particles is much higher than that of larger particles, and (iii) using an appropriate slot depth and blowing ratio increases the film cooling effectiveness and makes its distribution much more uniform, as well as reducing the deposition of graphite dust particles. The present findings are beneficial for designing direct-circulation helium turbines for high-temperature gas-cooled reactors.
Predictions of Component Remaining Useful Life Using Bayesian LSTM for Reactor Coolant Pump
Authors: Xu Renyi , Wang Yan , Peng Minjun , Cui Huaiming , Wang Hang , Kuang Chengxiao 
Published: November 2024

Abstract
With the increase of service life and coupling between components, reactor coolant pumps are faced with bearing damage, mechanical wear, shaft seal leakage and other performance degradation and safety problems. In order to predict the remaining useful life (RUL) of typical components of reactor coolant pump and meet the application requirements of predictive maintenance, a RUL prediction method for reactor coolant pump is proposed by combining Bayesian theory with data-driven method. In this work, we use Bayesian LSTM to predict RUL and its uncertainty, After clarifying the interface relationship of the prediction model, the model introduces arbitrary uncertainty and cognitive uncertainty into the deep learning network, and solves the problem of quantifying the uncertainty of the prediction results of data-driven method by constructing the RUL prediction model based on Bayesian LSTM. The research shows that compared with the neural network model with a single point estimate, the Bayesian LSTM can not only give the point estimate of the predicted result, but also provide the confidence interval of the RUL predicted result according to the data characteristics, thus providing more accurate technical support for the predictive maintenance of subsequent equipment. This has important prospect value for improving the status detection capability and intelligent operation and maintenance level of nuclear power equipment, and finally realizing advanced goals such as “no monitoring, few people on duty”.
Research on the Architecture of Unity3D-Based Rotating Equipment Fault Diagnosis System for Nuclear Power Plants
Authors: Jia Zhujun , Xia Hong , Yin Wenzhe , Jiang Yingying 
Published: November 2024

Abstract
This paper is based on the application of digital twin technology for nuclear power plant rotating equipment, aiming at relying on the virtual model to truly reflect the characteristics, performance and operation of the equipment to develop a fault diagnosis system architecture, which uses Unity3D as the carrier, combines MySQL and BP neural network fault diagnosis algorithm, realizes the communication and integration among the client, database and fault diagnosis algorithm. The fault diagnosis system is characterized by strong visualization, good interaction and high accuracy.
Advanced Radioactive Material Removal System Using Silver Zeolite (8) Development of High Temperature/Normal Pressure Test Equipment
Authors: Kawahara Yasuhiro , Narabayashi Tadashi , Endo Koji , Ishikawa Yoshihiro , Kobayashi Sanshiro 
Published: November 2024

Abstract
“Advanced radioactive material removal system using silver zeolite” aimed at developing an advanced filtered containment venting system or “FCVS” has been qualified for the 2020–2023 “Nuclear Industry Infrastructure Strengthening Project” by the Ministry of Economy, Trade and Industry. Advanced FCVS technologies, such as high-performance scrubbing nozzles, multi-layer metal fiber filters, and silver zeolite filters have been used to remove organic iodine. We have developed a new FCVS system, that has high-performance and is passively activated using SRV. This is believed to help eliminate clear infringement of personality rights, avoid litigation outages, and promote public acceptance for nuclear power to restart. The result of PRA analysis indicates that the loss of decay heat sequence called “TW” is reduced from 2.9 × 10−6 to 3.8 × 10−9. The total core damage frequency is reduced to almost 1/1000. The consortium to which the author belongs has conducted research aimed at improving the performance of these key technologies to enhance efficiency to remove radioactive materials and passive startup of the FCVS, and it has conducted evaluation tests on decontamination performance. These evaluation tests have mainly been conducted under the general atmospheric conditions of room temperature and atmospheric-pressure. In this study, with the aim of reproducing conditions similar to those in actual severe accidents, we have created a test device for high-temperature and high-pressure steam. This paper provides an overview of the test equipment.
Development and Validation of a Two-Phase Flow Solver With Drift-Flux Model Based on OpenFOAM
Authors: Wu Wenqiang , Zhang Dalin , Huang Tao , Zhou Lei , Du Peng , Tian Wenxi , Qiu Suizheng , Su Guanghui 
Published: November 2024

Abstract
The thermal-hydraulic system code is an important support of the design, safety and operation of nuclear reactor. With the rapid development of numerical methods and software engineering, the modernization of thermal-hydraulic system code development becomes more and more important. Based on the open source finite volume solution platform OpenFOAM, this paper develops a system-level two-phase flow solver for calculating reactor pipelines, aiming to realize the reactor thermal hydraulic system program through the idea of computational fluid dynamics. In this study, the drift partial non-equilibrium model is adopted to take into account the calculation accuracy and calculation speed. The governing equations of the two-phase flow model are converted into a form suitable for OpenFOAM programming specifications, and a series of constitutive model classes are established to close the governing equations. For the developed solver, other thermal-hydraulic system codes and experimental data are used to test and verify the solver’s ability to simulate single-phase, subcooled boiling, and flow boiling conditions. The results show that the solver developed in this paper successfully realizes the system-level two-phase flow calculation in pipelines. This work is the first step towards the ultimate goal of expanding the capabilities of the developed solver to the calculation of closed loop systems in the reactor.
Characteristics of Irradiation Production of Radioisotope in an Ultra-High Flux Research Reactor
Authors: Li Jian , Xu Wei , Zhao Jing , Liu Zhihong , She Ding , Xie Heng , Shi Lei 
Published: November 2024

Abstract
Radioisotopes are widely used in industry, agriculture, healthcare, and other fields, they are important basic materials for nuclear science and technology. The research reactor scheme designed by our institute is characterized by high neutron flux, flexible rector core layout, strong irradiation capability, and convenient operation. It has important applications in nuclear fuel and material irradiation test, radioisotope production, neutron science research, etc. By arranging irradiation targets in the thermal irradiation channels of this reactor to produce radioisotopes, it is expected to obtain satisfactory radioisotope yield. This paper analyzes the characteristics of radioisotope production with the high flux reactor, and the productions of rare isotopes, medical isotopes and industrial isotopes are considered. With the relatively high thermal neutron flux, the target irradiation period could be shortened and the specific activity would be improved. Based on the characteristics of radioisotope conversion, a reasonable arrangement scheme for irradiation targets will help to improve the neutron utilization and isotope production.
Study on the Flash Evaporation Characteristics of High-Temperature and High-Pressure Water Immersion Jet and the Behavioral Mechanism of Iodine in Liquid Pool
Authors: Wu Han , Gu Haifeng 
Published: November 2024

Abstract
A pressurized water reactor nuclear power plant’s coolant may experience a dramatic phase shift near the rupture point in the case of a Steam Generator Tube Rupture (SGTR) accident. This might cause steam to enter the secondary side or inject radioactive materials into the secondary side using a high-temperature, high-pressure jet. There could be flashing at the rupture location. The process of iodine-containing high-temperature, high-pressure water in an immersed jet during a rupture event is the main subject of this work. In order to provide a theoretical foundation for the examination of SGTR accident source terms, it examines the primary parameters impacting iodine’s evaporation, deposition, and mass transfer during the jet flashing process. Through a series of studies, it was seen and determined that the temperature and pressure of saturated water cause the flashing fraction to grow, with the upward tendency eventually slowing down. When high-temperature, high-pressure water jets undergo a specific level of supercooling, there is an approximately linear correlation between the flashing fraction and the degree of supercooling. Research on iodine-containing high-temperature, high-pressure water jets shows that the flashing process is not significantly affected by the iodine’s presence. The retention behavior of iodine in the liquid phase provides information for the analysis of SGTR accident scenarios and is dependent on a number of variables in the solution, including iodine concentration, supercooling, hydrogen potential, and others.