bluuebunny/crossref_metadata_2025 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

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

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

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

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

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

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

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

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

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

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

“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

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

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

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.
Dynamic Modeling for Compact Recuperator in Helium Gas Turbine Direct-Cycle System
Authors: Geng Xuyao , Wang Jie
Published: November 2024

The helium gas turbine direct-cycle scheme has the advantages of high operation temperature, simple design, and compact structure. It can serve as an efficient energy conversion system (ECS) and combine with various forms of energy, so it has broad prospects for development. Recuperator is crucial to helium gas turbine direct-cycle system. Its main function is taking advantage of high temperature helium at the outlet of turbine to heat microthermal helium behind the high pressure compressor to improve the power generation efficiency of the system. The accuracy and reusability of recuperator model will affect the system modeling and simulation. It is necessary to explore the appropriate modeling and solving methods for recuperator, and analysis the transient characteristics of recuperator under different operating conditions. In this paper, the plate heat exchanger is selected as the recuperator, and the segmented lumped parameter method is adopted to calculate the dynamic characteristics under the condition of mass flow rate and temperature disturbances. The calculation results are compared with those obtained by CFD methods. It indicates that the segmented lumped parameter model is robust and agree well with the results obtained by the finite volume method. The responses of recuperator to temperature and mass flow disturbance are analyzed. The flow rate has a greater influence on the metal wall temperature distribution shape. In power regulation, changing the flow rate synchronously on both sides of recuperator, the wall temperature is basically unchanged. In load rejection, if the regulating process lasts too long, the wall temperature may suffer from a cold shock.
Features of Space Reactor Brayton Cycle Under Power Regulation by Control Drums
Authors: Ma Wenkui , Ye Ping , Gao Yue , Hao Yadong , Yao Yi , Yang Xiaoyong
Published: November 2024

The development of space exploration technologies such as planetary bases and orbital propulsion has higher requirements for space power systems. The reactor coolant of the space reactor Brayton cycle (SRBC) is also the cycle working fluid, which has the advantages of high power conversion efficiency, compact configuration, small mass and volume. It is the one of the best options for future high-power space power systems. Reactivity insertion by control drums is an effective means of power regulation for the SRBC. In this paper, the dynamic model for the SRBC were established, the safety characteristics of the SRBC under power reduction conditions by control drum insertion were studied. The results indicated that the control drum insertion decreases the output power of the turbine and further leads to speed fluctuations. The bypass valve control can ensure stable shaft speed and balance between power generation and load of the generator, ensuring the safe operation of the SRBC under reactivity insertion conditions. Furthermore, the safety boundary for control drum insertion conditions of the SRBC was investigated. The results indicated that more control drum insertion decreased the output power of the turbine. The speed decreases far from the rated speed and exceeds the safety margin of the bypass valve control. The continuous decrease in speed leads to system shutdown, and there is a safety boundary. The analysis provides a reference for the operation and control of the SRBC in the future.
Depressurization Behavior and Strategy of Automatic Depressurization System for AP1000
Authors: Shang Zhengrun , Lv Dufeng , Meng Zhaoming , Sun Zhongning
Published: November 2024

This article through simulate small-break loss-of-coolant accidents (SBLOCA) specifically in the AP1000 and thoroughly analyze the behavior and strategy of the ADS in such accident scenarios. A comprehensive AP1000 model is constructed using the System Analysis Code, incorporating various components including the reactor coolant system (RCS), secondary system, passive safety injection system (PSIS), and automatic depresurization system (ADS). Five different types of conditions are simulated and evaluated to assess their thermal-hydraulic impact on the RCS, passive core cooling system, core degradation, and the ADS itself. The analysis results show that the design of two sets of independent ADS depressurization pipelines aims to achieve a redundancy level of 2 × 100%. Gradually opening the ADS pipelines increases the inlet pressure of the sprayer and widens its pressure curve, allowing for a sustained stable critical jet state over an extended duration. In the initial stage of the accident, opening the ADS stage with a large area stage can effectively improve the depressurization and cooling rate of the RCS and maintain the critical discharge state of ADS. ADS-4 can reduce the RCS pressure to the start-up pressure of In-containment refueling water storage tank (IRWST) to ensure long-term core cooling. These results hold significant meaning in comprehending the behavior and strategy of the ADS in typical SBLOCA scenarios, thereby contributing to the establishment of an effective theoretical foundation for the design of ADS pipelines and sprayers.
Numerical Investigation on Interphase Interaction and Interplume Effects of Multi-Hole Steam Injection Condensation Based on ADS1-3
Authors: Xu Jibin , Zheng Hao , Shen Xihao , Zhang Yuhao , Lu Daogang
Published: November 2024

Multi-hole jets are commonly employed in most technical applications for cooling and depressurization of high-temperature and high-pressure steam emissions due to their extraordinarily high heat and mass transfer efficiency. Due to the complexity of the micromechanics of steam condensation, in numerical calculations, existing models do not fully satisfy simulation needs, particularly under porous circumstances, past research has concentrated on single-pore steam jet direct contact condensation (DCC). Based on the ADS 1-3 level equipment in IRWST, the development of a model for DCC of multi-hole vapor injection, a three-dimensional transient numerical simulation research of the DCC mechanism and heat transfer characteristics of a multi-hole high-temperature and high-pressure steam jet is undertaken in this work.

The simulation results of the VOF and Mixture multiphase flow models graphically demonstrate the variations in their calculation theories, with the treatment of interphase interpenetration being the most significant difference. Compared to the MIXTURE model, the plume calculated using the VOF model is much larger, with the axial length of the plume being 3 to 4 times that calculated using the MIXTURE model, and the radial length being 2 to 3 times that calculated using the MIXTURE model, which is in gross disagreement with the experimental results. Finally, the Mixture model is better suited for numerical simulations of the steam jet DCC’s non-full Eulerian multiphase flow model. It serves as a data source for future porous steam jet DCC numerical simulation investigations.
The Mission Time Analysis in Level 2 Probabilistic Safety Assessment of the Third-Generation Nuclear Power Plant
Authors: Guo Shujie , Liu Jing , Zhong Yubao
Published: November 2024

Mission times have to be defined for the probabilistic system or equipment failure analysis. Severe accident mitigating systems may need to be in operation for a long time period, thus mission time considered in Level-2 probabilistic safety assessment (L2PSA) is usually longer than in Level-1 probabilistic safety assessment (LIPSA). By simplication, the mission times in the LIPSA are often defined globally as a maximized delay before obtaining a stable state without core damage. A common mission time is often defined as 24 hours for all systems and all accidents in LIPSA. L2PSA has another perspective. In practice, most LIPSAs with the issue of containment challenge due to core concrete interaction which may pressurize the containment or melt through the containment bottom within typical time scales of one to several days. Due to the complexity of the severe accidents and the diversification of the mitigation measures, there is no common time in L2PSA. Different times are used, like 24 hours or 72 hours. Selection principles of the mission time in L2PSA were researched and the modeling methods were developed. As an example of generation III NPP, the mission time was chosen and modeled, and the sensitivity was analyzed.
Numerical Simulation Research on Irradiation-Thermal-Mechanical Behavior of CDM Fuel in Different Types of Reactors
Authors: Tang Changbing , Li Yuanming , Liao Haoyu
Published: November 2024

The ceramic dispersion microencapsulated (CDM) fuels have high tolerance of accidents, and are regarded as promising candidates for the light water reactors. During irradiation, the parts in the CDM fuel may experience extreme irradiation-related irradiation-thermal-mechanical behavior. The buffer layer, which provide free volume for gaseous fission products and absorb kernel swelling, will firstly shrink during the early lifetime, and will re-contact to the IPyC layer. The re-contact of the two layers will cause high circle tensile stress in the IPyC layer, which may lead to the failure of the microencapsulated particle. In order to study the in-pile performance of the CDM fuels, and the possible application of the CDM fuels in other kind of reactors, such as the high flux reactor (HFR), the ADS reactor, and RERTR research and testing reactor, a series of researches are needed. Because of the economic efficiency of in-pile tests, researches of combined with numerical simulation could be an effective method for the study of the in-pile behaviors, the performance evaluation and the optimum design of the CDM fuel. In this research, a 3-D numerical simulation model of the in-pile behaviors of CDM fuel particles with ceramic matrix is built. Different irradiation conditions based on various type of reactors are introduced into the simulation of CDM fuel in-pile irradiation-thermal-mechanical behavior. The simulation results and the failure evaluation of the CDM fuels are discussed.
Modular Structure Selection and Design of ACP1000 Reactor Building
Authors: Xu Man , Tang Minghao , Jia Xiaopan
Published: November 2024

The Chinese nuclear power industry has entered a period of safe and efficient development, facing challenges such as an aging workforce, labor shortages, and significant energy and resource consumption in the construction process. Modular design and construction technology has been proven effective in the latest third-generation nuclear power projects, offering a means to reduce on-site construction, minimize safety hazards, shorten construction periods, and lower overall project costs. There are various structural forms of modular construction. This paper analyzes multiple modular construction technologies, examines the specific characteristics of the ACP1000 Pressurized Reactor, selects a structural modular type suitable for the ACP1000 Pressurized Reactor, and provides a detailed explanation of the modular design calculations for the internal structural buildings within the reactor facility.

Accelerating the implementation of modular internal structure design is of great significance for shortening the main construction period. The main body of the internal structure factory building is a shear wall and thick floor structure, which bear a large and complex load. The traditional reinforced concrete construction site has a large workload and long working time. Comprehensive analysis shows that the steel plate concrete structure is suitable for modular design of the internal structure. The steel plate concrete structure can meet the bearing and construction requirements of the internal structure due to its high construction efficiency and good stress performance. This article focuses on the design and calculation of steel plate concrete walls in practical engineering, and verifies them with reference to the “Technical Specification for Steel Plate Concrete Structures of Nuclear Power Plants “ GB/T 51340-2018, ultimately determining the design scheme of steel plate concrete structures.
The Influences of Key Factors on the Gas Diffusion in the Material Irradiation Test Based on the Regulation of Gas Composition
Authors: Sun Yixiong , Yang Wenhua , Sun Sheng , Zhang Liang , Zhao Wenbin , Jin Shuai , Lei Jin , Cheng Jinkang
Published: November 2024

Material irradiation test requires to design device structure to ensure that the specimen temperature can meet requirements. Gas composition regulation is one of the main temperature control methods. Temperature gradient is established through the gas gap between the specimens and coolant, and the thermal conductivity of gap can be changed by regulating the gas composition, thereby controlling the specimen temperature. The study takes the annular gas gap of a typical material irradiation device as the research object, adopts computational fluid dynamics (CFD) method to simulate the gas diffusion process, and analyzes the influences of key factors such as type of gas inflated, gas inflation rate and volume on gas diffusion behavior during the process of gas composition regulation. Helium floats because of its lower density than the mixed gas, and argon sinks because of its higher density. For the same working conditions, argon will diffuse to a farther location under the action of greater inertial force. The higher the gas inflation rate, the wider the range of gas diffusion to the surroundings at the bottom of the device, the faster the diffusion rate in the gas gap, and the longer the diffusion distance. The volume of only affects the final diffusion state. The higher the gas inflation volume, the mole fraction of the corresponding gas will continue to increase.
Data Sharing and Business Collaboration-the Refined Management Process of Bulk Materials for Nuclear Power Projects
Authors: Jie Wang , Zhuoze Li , Yuan Xin , Xinpei Xia , Yi Zhao , Sen Wang , Yanling Zheng
Published: November 2024

The bulk materials have the characteristics of many varieties, huge quantities, and frequent changes, which have made it difficult to manage the construction of nuclear power projects in the past, and the problem of high residual rate has occurred frequently. The traditional project management mode has the characteristics of focusing on the overall goal of the project and emphasizing the schedule over the cost, which leads to the lack of high requirements for the accuracy, standardization and timeliness of the quantity of bulk materials by the design, procurement and construction parties in the process of project implementation. Based on the principles of data sharing and business collaboration, this paper studies and constructs a set of data-based bulk material refinement management system process. The process is based on design data, compacts the business processes of on-demand procurement and on-demand picking, and loads procurement and construction information. In order to solve the data island phenomenon brought about by traditional form-based management and the problems of difficulty and error rate in manual processing of massive data, a series of software tools and management information systems have been developed to open up the data chain between the three sectors of design, procurement and construction, so as to realize real-time and accurate transparent information transmission.
Effect of Different Temperature Probability Curve Fitting Methods on the System Reliability for PCCS in AP1000
Authors: Yu Yu , Liu Guanyu , Feng Wanxin , Zhao Houjian , Lyu Xuefeng
Published: November 2024

Passive safety systems operating based on natural circulation are widely used in new generation reactor design to improve the safety of nuclear power plant, and functional failure of passive safety system gets more and more attention in recent years. Passive Containment Cooling System (PCCS) in AP1000 nuclear power plant is a typical passive safety system, and atmosphere is used as the heat sink, so air temperature is one of the important factors affecting the system operation. The air temperature probability density curve used in the system reliability evaluation is always gained based on the historical data through fitting ways such as least square method. However, the system always has large safety margin, so for the system failure cases, the air temperatures are so high that have quite low probability. And these values are in the tail of the temperature probability density curve, so the fitting methods may have important effect on the results. In this paper, we calculate the system failure probabilities for PCCS in AP1000 with two temperature probability curves gotten from different fitting ways, and the effect of the air temperature probability curve fitting method is analyzed.
CFD Modeling of a Full Clogging Event in a Pumping Station With Neptune_cfd Solver: A Preliminary Investigation
Authors: Natchez Sebastien , Ceyrolle Romain , Loizeau Vincent , Cordier Hervé
Published: November 2024

The drum screen filters the water upstream of the pumping station of a NPP to protect it from debris entrance. These devices can be damaged in case of full clogging events because of drum deformation or air aspiration in the pumps located downstream of the filter. With climate change and the contingency to face more clogging events, EDF has lately tried to estimate their impact on pumping stations by different manners. The focus was set on passive filtration (static filter with no cleaning system) to reduce the complexity of the modelling and the objective is to determine how long it would take for the water level downstream of the filter to reach a hazardous level. To study such cases numerically, it is necessary to select the appropriate physical models which are able to simulate the physical phenomena while minimizing code developments. In this context, a first attempt with code_saturne highlighted the necessity of a two-phase flow code to simulate an industrial application, which was confirmed by a second attempt with neptune_cfd. In order to ensure the reliability of the results it is now required to validate the code on simpler problems where experimental data are available. This preliminary investigation shows that a development of the deposition model is still necessary to use a lagrangian method for particle transport and deposit in neptune_cfd.
Experimental Study on the Coolability of Molten Core Materials Discharged Into a Depth- and Volume-Limited Sodium Plenum
Authors: Matsuba Ken-ichi , Kato Shinya , Kamiyama Kenji , Akaev Assan S. , Vurim Alexandr D. , Baklanov Viktor V.
Published: November 2024

During a severe accident in sodium-cooled fast reactors, molten core materials could be discharged from the core region toward the lower sodium region of the reactor vessel through coolant channels, such as control rod guide tubes. Typical SFRs have a sodium plenum with limited depth and volume, such as the core inlet plenum located under the core region. Therefore, it is important to evaluate the coolability of molten core materials discharged into a depth- and volume-limited sodium plenum. In the present study, to deepen the understanding on the coolability of molten core materials discharged into such a sodium plenum, conditions under which molten core materials form solidified fragments were discussed based on an experiment discharging a molten fuel simulant (molten Al2O3) into a test vessel filled with liquid sodium. The data obtained in this experiment suggest that the molten Al2O3 formed solidified fragments under a condition beyond the cooling limit (the volume ratio of sodium to molten core materials: more than 10 times) estimated from existing studies, such as M-series experiments performed at the Argonne National Laboratory in the 1970s. Sodium convection flow induced by sodium vaporization (vapor bubble growth) and condensation during the Al2O3 discharge might have contributed to such efficient cooling.
Proposal for Organizing and Understanding Decommissioning Information Using Task Ontology
Authors: Taruta Yasuyoshi , Iguchi Yukihiro , Kawasaki Daisuke , Yanagihara Satoshi , Tomoda Koichi
Published: November 2024

In this study, we propose a hierarchical framework using task ontology, which is used in information science and information engineering, to organize decommissioning information. As background, decommissioning of nuclear facilities generally takes a long time, so it is important to accumulate information and utilize it.

In decommissioning, the operator creates a plan using facility information during construction, subsequent renovations, and contamination distribution and evaluation during operation, and conducts decontamination, demolition, waste management, and waste delivery. It may also include restoring the land on which it was installed to its original condition. In addition, there are several implementation policies for this decommissioning: cases where the operating entity hands over the decommissioning to the decommissioning company, and cases where the operating entity directly carries out the decommissioning. Decommissioning is being carried out in each country and by various organizations, and various knowledge and lessons are being created every day. It is important to utilize these findings. On the other hand, it can be said that it is difficult to simply utilize it in other organizations due to different laws and regulations and different organizational management policies. It is necessary to appropriately utilize precedent cases and information on lessons learned.

Therefore, in this study, we introduce the concept of ontology to decommissioning and propose a conceptual model from the perspective of plant information management and knowledge management. Ontology has guidelines for understanding objects without using language, and is a conceptual structure of knowledge. Previous studies and reports have referred to the importance of ontology in decommissioning, but it cannot be said that its structure and relationships with other ontologies have been specified.

In this study, we propose a structural organization using task ontology as a guideline for information organization to cope with such situations.
Heat Transfer to Supercritical Water Flowing in Short Vertical Bare Tubes
Authors: Kavalci Mehmet , Wspanialy Mark , Heyns Laura , Cornelius Marcus , Pioro Igor
Published: November 2024

It is well known that water-cooled nuclear-power reactors / Nuclear Power Plants (NPPs) equipped with light- and heavy-water-cooled reactors have relatively low thermal efficiencies within the range of 26–38% compared to those of thermal power plants – combined-cycle up to 62.5% and supercritical-pressure Rankine cycle up to 55%. Therefore, increasing thermal efficiency of next generation NPPs / nuclear reactors is a very important task. For water-cooled reactors / NPPs the conventional approach for increasing thermal efficiency is to move to supercritical pressures, as it was done more than 60 years ago in the thermal-power industry.

Currently, together with large-capacity SuperCritical Water-cooled Reactors (SCWRs) SC-Pressure Small Modular Reactors (SCP SMRs) are also being developed. Therefore, solid knowledge of heat-transfer specifics at supercritical pressures is required. Usually, hydraulic-equivalent diameters of modern bundle flow geometries are within 5–10 mm and heated lengths are up to 4 m for SCWRs and can be shorter for SCP SMRs.

Also, first SCWRs as test, experimental, or pilot reactors will be with relatively small installed capacities within 100–300 MWel. Due to that it is important to investigate supercritical heat transfer in small diameters tubes as a first and preliminary step instead of bundle geometries and bundle simulators. Usually, heat transfer in bare tubes is considered as the conservative approach compared to bundle tests, because heat transfer in bundle geometries will be higher compared to that in bare tubes due to additional turbulization from various appendages inside sub-channels.

The paper presents experimental data obtained in relatively short vertical bare tubes mainly with upward flow of SCW. The experiments have been performed by Dr. V. Razumovskiy1 at the National Technical University of Ukraine. The data were obtained at the pressure of 23.5 MPa, mass fluxes of 2193 kg/m2s (main set of data) and 509 and 248 kg/m2s, and heat fluxes within 104–2660 kW/m2.

The experimental data were compared to a well-known subcritical-pressure heat-transfer correlation developed by Dittus and Boelter. The heat-transfer-coefficient data were predicted quite reasonably with this correlation within the normal heat-transfer regime, but were significantly higher those within the deteriorated heat-transfer regime and within the pseudocritical region.