bluuebunny/crossref_metadata_2025_split huggingface dataset:
Research on the Criteria for Single Protection Channel Power Failure in Nuclear Power Plants
Authors: Zhang Xiuchun , Xia Hong , Liu Yongkang , Zhu Shaomin , Zhang Jiyu , Xu Junjun , Jiang Yingying , Yin Wenzhe , Zhang Jie , Liu Jie 
Published: November 2024

Abstract
In order to ensure the safe operation of a single protection channel in a nuclear power plant in case of power failure, it is necessary to carry out the consequence analysis of power failure in the construction drawing design stage. Moreover, in order to ensure the effective application of engineering design experience in a new power plant, it is necessary to establish design criteria and operation management criteria for the power failure of a single protection channel. Based on the principle of digital design of protection channel in a typical nuclear power plant, this paper analyzes the behavior, main consequences, possible safety injection risks, precautions for power transmission restoration after power failure of a single protection channel, and explains the verification process of power failure analysis results of a single protection channel in the design process, the design idea of corresponding operation documents and the logic of the operator’s execution of operation documents. Through the analysis and research process, the design criteria and operation criteria applicable to the power failure of single protection channel of nuclear power plant are obtained. The research results of this article can provide a basis for the design and operation management of single protection channel power failure in newly built nuclear power plants in China. Furthermore, it ensures the operational safety of nuclear power plants in the event of a single protection channel failure.
Research and Application of Containment Structural Health Monitoring System During Operation of Nuclear Power Plant
Authors: Mengyan Song , Di Yao , Chaoqi Zhang , Di Jiang , Peiyao Zhang 
Published: November 2024

Abstract
The containment of the nuclear power plant is an important item of the tolerance nuclear steam supply system. It is the third barrier to prevent radioactive waste from leakage after an accident. It is the most important structure of nuclear power plant ensuring the safety of nuclear power plants. To ensure that the containment meets the design requirements during the service period during service, Structure integrity test is done every ten years to verify whether the performance of the containment meets the design requirements. However, through the cycle Structure integrity test, the structural state of the containment is obtained every ten years. It is unable to obtained in real time. When the accident conditions occur, the structural state of the containment cannot be evaluated quickly and unable to assist the containment related decision -making. In order to ensure the safe operation of the containment during the service time, at the same time improve the intelligent level of the safety monitoring of the containment, developing a set of intelligent monitoring and evaluation and early warning systems to achieve daily monitoring, abnormal state warning, healthy real-time real-time monitoring, abnormal state warning, abnormal state warning, abnormal state early warning, healthy Status assessment are very necessary.
Study on Influence of Braking Circuit and Stator-Rotor Structure on the Damping Characteristics of Permanent Magnet Motor
Authors: Ren Zhiyuan , Wu Hongyu , Yan He , Li Tianjin , Diao Xingzhong 
Published: November 2024

Abstract
Permanent magnet motor is used as the power source to provide driving force for the lowering, lifting and holding action of the control rod in high-temperature gas-cooled reactor. When reactor scram and the rods falling by gravity, the permanent magnet motor also can act as a damping system to convert the gravitational potential energy of the control rods into the thermal energy consumed in the braking resistor, that enables a damping function of limiting the speed along the long dropping stroke and effectively buffering the impact of the control rods on the reactor internals. In order to further investigate the damping capability of the motor, a finite element model was established for the permanent magnet motor, as well as its external circuit. The braking and damping performance of the motor under different rod drop speeds was analyzed by simulation methods. An important performance parameter for evaluating the damping capability of motor is the damping torque, which is mainly affected by the braking resistance in the external circuit model and the stator-rotor structure of the motor. First, a full-size motor with external circuit model was established based on Ansys Maxwell software, and reasonable meshing and irrelevance verification of the motor body model was carried out to improve the solution accuracy and modeling efficiency of finite element simulation. Then the influence of the motor structural parameters (such as air gap, stator-rotor tooth shape, permanent magnet structure, and braking resistance setting of the external circuit) on the damping torque was studied through parameter-sensitive simulation and analysis, especially through the reasonable setting of the external circuit the adjustment of the damping torque was realized. The effect of motor structural parameters on damping performance was also obtained. According to the required falling speed of the control rod, the appropriate resistance value could be selected to regulate the falling rod speed, which provided a strong support for the realization of the falling rod buffer of the control rod drive mechanism with the permanent magnet motor serving as the second damping system. At present, the research work had been carried out based on the finite element model of the permanent magnet motor, which would provide a validation basis for the subsequent experimental research.
Simulation Study of Nuclear Heating Reactor Control Method in Slight Boiling Operation Mode
Authors: Huang Tianyi , Sun Jun , Xie Heng 
Published: November 2024

Abstract
Nuclear Heating Reactors (NHRs) could achieve multi-purpose target including heating and seawater desalination without emission of greenhouse gases. Institute of Nuclear and new Energy Technology (INET) have designed two operation modes for NHRs: pressurized mode and slight boiling mode. The former shows a higher level of maturity, while the latter guarantees better economic outcomes but still lacks comprehensive testing and analysis. Previous researches on operation of NHR slight boiling mode mainly concentrate on avoiding instabilities during reactor start-up process, while the power adjustment process is also important but seldom researched. NHR-200II/S is designed to obtain the slight boiling operation capability based on the parameters of NHR-200II, and RELAP5 code is employed for the simulation of operation processes. A proportional-integral-derivative (PID) control scheme is introduced to maintain the primary loop pressure by adjusting the flow rate of intermediate loop. The PID coefficients are determined and the effect of the control scheme is verified by power adjustment processes. The ability of proposed control scheme to maintain stable operation of NHRs in slight boiling mode is shown.
Concept Study of a Fission Fragment Propulsion System
Authors: Zhang Dacai , Yu Ganglin 
Published: November 2024

Abstract
Deep space exploration holds immense significance for the future of space exploration, and it requires propulsion systems with high specific impulse characteristics to accomplish long-duration space missions. The performance of prevalent nuclear thermal and nuclear electric propulsion systems is constrained by fuel element cladding temperatures and energy conversion systems efficiencies, precluding further enhancement of specific impulses. This paper focuses on a space nuclear propulsion system with a magnetic collimation structure based on the direct propulsion of fission fragments of Am-242m. The system employs eight control rods for reactive adjustments, the system’s critical and propulsion characteristics are analyzed by the combination of RMC and MATLAB. The calculation results demonstrate that the system exhibits superior comprehensive performance, with a specific impulse of approximately 4 × 104s while the keff of the system is optimized to reach about 1.06. The results also indicate that the system possesses excellent safety features and can play a crucial role in deep space exploration. The exploration of propulsion systems utilizing Pu-239 and U-235 as well as the investigation of enhancing the fission fragment escape rate and energy utilization rate will be continued in the future.
Numerical Study of Flow-Induced Vibration of Square Tube Bundle Structures Under Ocean Conditions Based on Large Eddy Simulation
Authors: Sun Junshuai , Sun Rulei , He Xiaoqiang , Wen Jiming , Tan Sichao , Tian Ruifeng 
Published: November 2024

Abstract
Marine nuclear power equipment is an important support to solve the problems of long-lasting power energy supply in the development of far-reaching marine resources and “carbon emission reduction” in the marine field. The additional force under the ocean environment will lead to the change of the flow field characteristics of the tube bundle structures of the steam generator in the nuclear power equipment, which in consequence will lead to the change of the flow-induced vibration characteristics, and the safety of the system will be tested. In order to study the flow-induced vibration characteristics of the tube bundle structures under ocean conditions, taking the square tube bundle structure as the research object, based on the large vortex simulation method, using the secondary development of customized functions, combined with the local dynamic mesh deformation technology, the fluid-structure interaction computational model of the tube bundle structures under ocean conditions was established. Firstly, the effect of grid discretization on the predictive ability of the flow field characteristics of the tube bundle structures was studied by the calculation of the flow field of the square tube bundle structures under the static conditions with typical parameters. Secondly, the accuracy of the fluid-structure interaction computational model was verified by the calculation of the flow-induced vibration of the square tube bundle structures under the non-ocean conditions. Thirdly, the flow field characteristics of the square tube bundle structures under the ocean conditions was researched and compared with the results of the non-ocean conditions, the effect of oceanic conditions on the flow field characteristics of the tube bundle structure was obtained. Finally, the flow vibration characteristics of the square tube bundle structures under ocean conditions were investigated and compared with the results of the non-oceanic conditions, the effect of ocean conditions on the flow-induced vibration characteristics of the tube bundle structures was obtained. The results show that the established fluid-structure interaction computational model can accurately predict the flow-induced vibration behavior of the tube bundle structures under oceanic conditions; the existence of ocean conditions will lead to the generation of transverse flow in the flow field of the tube bundle structures, which will enhance the fluid mixing between the tube bundles and make the flow-induced vibration more severe. This study is of great significance for the design and optimization of steam generator tube bundle structures in the marine nuclear power equipment.
Pool Boiling Heat Transfer Enhancement Using Bi-Conductive Surfaces
Authors: Tang Longchang , Liu Xiaojing , Xu Wei , Tang Luyao 
Published: November 2024

Abstract
Due to its superior heat dissipation capability, pool boiling has been a research hotspot over the decades. This paper provides insights into the underlying pool boiling enhancement mechanism of bi-conductive surface which is an effective modification method. In this study, a new type of bi-conductive surface is created by embedding a reticular structure made of low-conductive epoxy into a high-conductive copper substrate. The influence of the width of epoxy is studied by conducting experiment as well as CFD simulation. Due to extreme differences in thermal conductivity, the heat flux almost exclusively flows throw copper. Moreover, epoxy inhibits nucleation because of its low surface energy. Bubbles are almost only generated on copper and the colder epoxy remains moist. The existence of epoxy creates periodic spatial surface temperature variations and induces separated liquid and vapor paths. As a result, compared with bare copper surface, bi-conductive surfaces have better heat transfer performance. For the bi-conductive surface with 0.75mm width of epoxy, replacing 44% of the high-conductive copper with low-conductive epoxy results in 270% increase in CHF value and 108% increase in the total effective thermal power at CHF.
Experimental Study on Characteristics of Entrained Droplets and Liquid Film at Swirler Outlet in Swirl-Vane Separator
Authors: Kang Ruiqi , Xiong Zhenqin 
Published: November 2024

Abstract
The swirl-vane separator, recognized for its compact design and high separation efficiency, experiences a gas-liquid flow transition from non-swirling to swirling after passing through the swirler. Driven by centrifugal force due to gas-liquid density disparities, efficient separation is achieved as liquid is propelled towards the wall. However, complexities arise in the flow filed within the separator due to significant variations in cross-sectional area from upstream to downstream of the swirler. In conditions of insufficient gas flow or low liquid phase flow, a swirling churn flow emerges at the swirler outlet, causing fluctuations in the liquid film and entrained droplet generation. The limited separator riser height impedes timely droplet-wall contact, leading to compromised separation efficiency. To delve into this complexity, an air-water experimental facility was established, investigating the two-phase flow at the swirler outlet. Wire mesh sensors gauged liquid film and droplet characteristics under varying conditions (air superficial velocity: 12.4–20.4 m/s, liquid superficial velocity: 0.1–1.3 m/s). Results revealed that liquid film thickness increases with decreasing gas velocity and rising liquid velocity. Entrained droplet diameter correlates positively with liquid velocity and swirl vane inclination but inversely with gas velocity. Notably, a 45° swirl vane inclination induced dispersed liquid droplets within the swirling annular gas core. Understanding these distribution characteristics is pivotal for setting separator working conditions to enhance separation efficiency and optimize swirler structure.
Experimental Investigation of Early Response of Hot Water Rapid Blowdown in Separated Pipeline
Authors: He Xiaoqiang , Gao Puzhen , Zhang Ze , Wang Jianjun 
Published: November 2024

Abstract
In this study, a series of separate effects tests are carried out to investigate the phenomena of the early response of the blowdown process in the ruptured pipe. The experimental section is a 50 mm nominal stainless steel pipe about 1600 mm long. Several high-precision piezoelectric sensors are placed along the pipe and the outside space near the rupture point with a data acquisition frequency of up to 20 kHz. Two methods are used for the postulated pipe rupture simulation: one using a quick-opening solenoid valve, and the other is realized by a thin glass cap and shatter trigger. Based on the experiments, pressure distribution and its time trends are obtained during the rapid blowdown process with various initial conditions, including different pressures and subcooling degrees. The characteristics of different stages in depressurization are identified, and some other vital parameters, such as the rate of depressurization and the initial discharge speed are also analyzed. In addition, the time response curve of back pressure in the outside space is recorded, and the flashing of superheated water during the blowdown is filmed. Moreover, the initial speed of the discharge flow is calculated after the data processing. The results of the tests help to understand the transient blowdown process and provide verified data for the simulations. It also provides a foundation for future experiments under higher working conditions and for developing boundary models used in analysis code.
Study on Heat Transfer Characteristics of 2×2 Helical Cruciform Fuel Assembly Under Non-Uniform Heat Source
Authors: Luo Menghao , Lu Zhiwei , He Yanan , Wu Yingwei , Tian Wenxi , Su Guanghui , Qiu SuiZheng 
Published: November 2024

Abstract
Helical cruciform fuel, due to its unique geometric shape, affects heat transfer, thereby improving the power level and safety margin of existing pressurized water reactors. This article establishes a unidirectional neutron thermal coupling calculation method for the complex geometric cross-section of helical cruciform fuel to study its flow and heat transfer characteristics under non-uniform heat sources. Based on the Monte-Carlo particle transport encoding OpenMC, the fission power distribution of the two-dimensional cross-section of a helical cruciform fuel rod were studied, and we used this as a non-uniform heat source for calculation of flow and heat transfer of 2 × 2 fuel assembly. The neutron calculation results show that the fission energy is concentrated at the fuel petal convexity, and the radial power factor peak fluctuates around 1.3. The flow heat transfer calculation results show that the temperature and heat flux at the petal convexity of the fuel rod cladding outer are lower than those at the petal concavity due to the longer heat conduction path; the swirling flow in the coolant channel enhances the heat transfer on both sides of the petal convexity, and flattens the heat flow distribution at the petal convexity, resulting in a smaller heat flow peak here. The trend about distribution of heat flow and wall temperature on the outer surface of the cladding under non-uniform heat source is basically the same as that under uniform heat source, with a relative deviation of 2.38% for temperature and 18.32% for heat flow. The influence of large heat flux deviation on the local thermal design of fuel cannot be ignored, and it is necessary to further optimize the coupling method and accurately the actual deviation value.
Analysis of Turbulent Structures Based on Large Eddy Simulation in Rod Bundle Channels With Mixing Vanes
Authors: Zhang Lixuan , Chen Guangliang , Tian Zhaofei , Qian Hao , Zhang Yuhang , Li Rui 
Published: November 2024

Abstract
The refinement of thermal-hydraulic calculations for reactors is crucial for both the economy and safety of reactors. However, the spacer grids and mixing vanes will affect the turbulence and vortices. In particular, the prediction of near-wall turbulence using computational fluid dynamics (CFD) is more difficult due to the strong shear stress and anisotropic properties near the coolant channel walls. By studying the turbulent coherent structures in the core coolant channel, it is possible to better predict the turbulent flow and obtain more accurate core thermal-hydraulic parameters. This paper focuses on the experimental objects of AFA-2G spacer grids and mixing vanes and establishes a model that meets the requirements of Large Eddy Simulation (LES) to study turbulent structures. The parameters of the flow field acquired from LES were in good agreement with the experimental results. By comparing with the Reynolds time-averaged model, it was found that its mean field is more accurate. This paper quantitatively analyzed the anisotropic characteristics of coherent structures and visualizes their spatial distribution using barycentric map. The visualization of vortices was performed using Ω criterion, and the dynamic analysis of vortices was conducted to investigate the effects of spacer girds and mixing vanes on the generation and evolution of vortices, as well as their mutual interactions.
Study on the Spent Fuel Management in Different Developed Countries
Authors: Shi Lei , Su Jiqiang , Chen Ding , An Yan 
Published: November 2024

Abstract
Spent fuel is nuclear fuel that has been exposed to radiation, usually produced by nuclear reactors in nuclear power plants, and is an inevitable product of the development of nuclear energy. Spent fuel contains a large amount of radioactive elements, which are characterized by high levels of radioactivity, difficult processing techniques, and strict safety management requirements. It is related to the sustainable development of nuclear energy, as well as the safety of the environment and human life. Ensuring long-term safety management of spent fuel has always been an important prerequisite for countries around the world to develop nuclear energy. To this end, major nuclear power countries around the world have established complete management systems and mechanisms, formulated sound policies and regulations, and ensured the systematization, standardization, and effectiveness of the entire process of spent fuel management. This study focus on national management policies and regulatory systems, investigates the practices of spent fuel management in major nuclear power countries such as France, Japan, and South Korea, and summarizes experience to provide relevant inspiration and suggestions for the management of spent fuel in China. Study has shown that it is necessary to strengthen top-level design, clarify the responsibility and authority boundaries of all parties involved, and to formulate policy systems such as the long-term development plans for the industry.