bluuebunny/crossref_metadata_2025_split huggingface dataset:
Analysis of Behavior of Prestressed Containment Under Close-in Explosion Based on CEL Method
Authors: Li Rongpeng , Yao Qinqin , Liu Mengsha , Jiang Di 
Published: November 2022

Abstract
The prestressed containment structure is the last physical barrier against the spread of radioactive material. In addition to being able to resist the pressure caused by internal accidents, it should also have a certain ability to resist external projectiles or explosive shocks. Under the background of frequent terrorist incidents and intensified local conflicts, evaluating the performance of the containment structure under the close-in explosion of explosives carried by drones or light weapons is of great significance for maintaining the safety of nuclear power plants. In this paper, the CEL (Coupled Eulerian and Lagrangian) method is used to calculate the overpressure of the air blast, which is compared with the empirical formula. The test of the damage to the concrete slab under TNT near-in explosion reported in the literature is simulated, and the test results are compared. A numerical model of containment was established and calculated to predict damage behavior. The results show that the air overpressure simulated by the CEL method is in good agreement with the empirical formula, and the simulated results are consistent with the experimental results reported in the literature. Finally, predictions are made for the performance of the containment model under the close-in explosion. This method can be used to design or evaluate the performance of containment against close-in explosions.
Measurement of Two-Phase Flow: Static CT System Based on Carbon Nanotubes
Authors: Zhang Yucheng , Xu Shuo , Xiang Xincheng 
Published: November 2022

Abstract
Two-phase flow measurement with high precision plays a significant role in safe and efficient operation of nuclear reactor. This paper focuses on the void fraction of gas-liquid flow in a non-transparent tube with the inner diameter of 13mm. X-ray computer tomography (CT) is considered to be effective for two-phase flow measurement because of its good qualities of non-invasion. However, due to the mechanical limitation of rotating speed, solutions for fast CT system to reduce the scanning time have been suggested relied on multiple sources and detectors. Recently, carbon nanotube (CNT) based X-ray source has been applied to CT imaging, significantly improves the temporal resolution by increasing the number of sources and avoid the gantry rotation. This paper proposes a potential static CT system design for the imaging of two-phase flow in straight steel tube. The setup of this system employed 90 couples of CNT X-ray sources and detector arrays arranged in a circle. Gas-liquid flow was simulated with different sizes of spheroidic bubbles randomly placed in the water inside the tube. To substitute for the flow moving, the z-axis of the simulator was added according to the flow velocity and exposure duration. Iterative image reconstruction was applied for the inverse problem of density distribution, and the reconstruction result of the experiment indicates the static CT system is useful to distinguish the flow pattern and measure the void faction distribution of gas-liquid flow.
Research on Precision Water Reactor Fuel Servo Operation System
Authors: Ping Li , Deng Peiqiang , Qian Wei , Yang Kaiyu , Zhai Jiansong , Huang Shuiyang 
Published: November 2022

Abstract
The PMC system of PWR nuclear power plant is a servo control system based on PMSM permanent magnet synchronous motor. The traditional Clarke and park coordinate transformation theory algorithm is used for precise control. This paper analyzes the composition of the servo system of PMC loading and offloading equipment, the principle and control algorithm of PMSM permanent magnet synchronous motor, and studies the double closed-loop control model based on current sampling and position sampling, the protection function setting of servo system and the self-learning method of PMSM. The research results show that the PMC system adopts the double closed-loop control theory of field oriented control (FOC) with DSP as the control core. Through the close-loop cooperation of PLC, servo driver, PMSM permanent magnet synchronous motor and resolver, the precise control of speed loop and current loop is realized, and the three closed-loop is realized by the position loop of PLC Control. This study provides a reference for the realization of independent deep maintenance of PMC equipment, and provides theoretical support for the future transformation of PMC servo control system and fault prediction of servo system.
Optimal Design of Thermal Scheme for Liquid Metal Cooled High Flux Reactor Fuel Assembly
Authors: Pan Rui , Zhang Kefan , Zhang Xilin , Deng Jian , Zhang Yong , Zhu Dahuan , Chen Hongli 
Published: November 2022

Abstract
This paper studies the structural parameters of fuel assemblies suitable for high neutron flux density environments.
Due to the high neutron flux density in the core, high-flux reactors provide an experimental environment for neutron irradiation. However, high neutron flux leads to high heat flux density on the fuel assembly surface, which brings challenges to the design of fuel assembly. Therefore, it is very important to study the structural design of fuel assemblies suitable for high neutron flux density environments.
Through theoretical derivation of the thermal model of the fuel assembly and sensitivity analysis of the design parameters of the fuel assembly using a single-channel program, the results show that the liquid metal cooling plate fuel element can be well adapted to the high neutron flux Density environment; at relatively low neutron flux densities, bundle fuel elements can also meet reactor design requirements.
Comparison Of Radiation Embrittlement Prediction Models for Reactor Pressure Vessel
Authors: Cai Kexin , Huang Ping , Qian Wangjie , Fan Minyu , Qi Shuang , Zhang Guodong , Zhang Yanwei 
Published: November 2022

Abstract
In order to promote the establishment of the material prediction model of domestic reactor pressure vessel (RPV) under high neutron flux irradiation, the prediction model of reference toughness and brittle-transition temperature variation (ΔTR,NDT) in foreign relevant guidelines and standards were compared and analyzed from the aspects of data basis, scope of applicable neutron flux and reliability of results. The results show that all prediction models are based on a large number of experimental data, and their reliability varies greatly. RG1.99 (Rev.2) has the worst reliability, while FIS has the highest reliability. When the neutron flux is less than or equal to 7.2 × 1019 cm−2 (E ≥ 1.0 MeV), the results of JEAC 4201 model are more conservative than NUREG/CR-6551, but when the neutron flux is larger than 7.2 × 1019 cm−2 (E ≥ 1.0 MeV), the results are opposite. The range of neutron flux applicable to each model is different, among which RG1.99 (Rev.2) and JEAC-4201 prediction models are the highest, reaching 10.0 × 1019cm−2 (neutron energy E ≥ 1.0 MeV). The establishment of the prediction model of domestic RPV materials under the condition of high neutron flux irradiation requires a large number of irradiation tests and in-depth research on the irradiation embrittlement mechanism, which lays the data and theoretical foundation for the establishment of the prediction model of domestic RPV materials under the condition of high neutron flux irradiation.
Study on Flow and Heat Transfer Performance of Space Reactor Radiator Water Heat Pipe
Authors: Jiao Guanghui , Xia Genglei , Tao Zhou , Bi Yuepeng 
Published: November 2022

Abstract
To ensure that the heat of the space reactor can be exported in time, the parameters and operating characteristics of the radiator heat pipe must be clear, and the heat pipe must work under various heat transfer limits and meet the safety requirements of avoiding single-point failure. Based on the basic principle of the heat pipe and fully considering the phase change characteristics of the gas-liquid interface in the heat pipe, this paper establishes the numerical calculation model of the operation characteristics of the space reactor radiator water heat pipe. The influence of phase transition on flow and heat transfer is considered. Through the calculation of the heat pipe, the operation parameters such as temperature, pressure and speed of heat pipe are obtained. On this basis, according to the requirements of safety characteristics analysis in the heat pipe, four heat transfer limit calculation methods of heat pipe viscosity limit, sonic limit, entrainment limit, and capillary limit are established. The heat transfer limits of the heat pipe of the space reactor radiator are calculated by using the constructed model, and the main heat transfer limits affecting the heat transfer performance of the heat pipe are determined. By coupling the heat pipe calculation program with the heat sink calculation model, the temperature distribution of the radiator panel is obtained, and the influence of the length of the radiator panel on the heat transfer characteristics is analyzed.
Research on the Reactor and Turbine Matching Key Technology of CPR1000 Nuclear Power Plant for Industrial Steam Supply
Authors: Mingjia Zhou , Kai Wang , Wei Zhang , Jianmin Zhu , Yulong Mao 
Published: November 2022

Abstract
The main steam extracted from the secondary side of nuclear power plant for steam supply in industry will affect the power matching between reactor and steam turbine, and the adaptability of reactor control system is the key of reactor-turbine matching research. By analyzing the influence of main steam extraction on reactor control system in CPR1000 nuclear power plant, a new reactor-turbine matching signal is designed and verified by modeling and maximum extraction steam flow rate is discussed. This study can provide important reference for the development of other nuclear energy comprehensive utilization.
Simulation of One-Dimensional Nuclear-Pumped Laser Generation System Based on Modelica Language
Authors: Tang Xiang , Wang Xiang 
Published: November 2022

Abstract
Nuclear-pumped lasers (NPL) utilize nuclear energy pumping to generate laser light and represent a viable approach to directly converting nuclear energy into useful products (coherent photons). Most nuclear-pumped lasers use thermal neutrons provided by pulsed nuclear reactors to initiate nuclear reactions. Nuclear-pumped lasers that use nuclear reactors as thermal neutron sources are also known as reactor-pumped lasers (RPL). Gas lasers mainly use inert gas as the laser medium, which has the characteristics of high output energy and long service life. This paper mainly establishes a one-dimensional He-Xe laser system, which can excite a laser with a wavelength of 2.03 μm, and the efficiency can reach 1% to 4%. According to the different physical processes of the laser generation system, we divided the system into three main components: nuclear reaction component, pump reaction component, and laser generation component. Based on the modeling method of Modelica for physical process componentization, it is used to simulate and verify each component, and the relevant parameters of the component are determined through theoretical calculation and numerical simulation. The componentized modeling method increases the universality and ease of editing of the model, and facilitates the maintenance and later expansion of the model. When the neutron fluence rate is 6.0 × 1014 cm−2·s−1, the laser with a wavelength of 2.03 μm and a power of 38.55 W can be output.
Structural Design and Dynamic Analysis of the Heat Exchanger Used in Nuclear Fuel Test Loop
Authors: Gang Huang , Sheng Sun , Liangming Pan , Junping Si 
Published: November 2022

Abstract
The heat exchanger was required to cool the high temperature and high pressure primary water from the test chamber of the nuclear fuel test loop generated by test fuel assembly. The heat exchanger was composed of shell-tube heat exchanger group and double-pipe heat exchanger group, the regenerative structure was adopted to greatly reduce the thermal stress caused by the large temperature difference between primary water and secondary water, and the long-term safe and stable operation of the device was ensured, the establishment of 3D analysis model and dynamic analysis of the heat exchanger were carried out in the paper, stress strength of the heat exchanger by different loads under design condition, normal condition, abnormal condition, accident condition and test condition were obtained, stress strength assessment for the heat exchanger structure based on RCC-M specification design and construction rules for mechanical components of PWR nuclear island was also carried out, the results indicate that the strength of the heat exchanger satisfied requirements of every RCC-M level.
Neutronic and Thermal-Hydraulic Analyses for an Additive Manufacturing Reactor Core With SiC Matrix and TRISO Particle Fuel
Authors: Han Wenbin , Deng Jian , Lu Qi , Chen Chong , Xu Youyou , Zhang Tao , Huang Shanfang 
Published: November 2022

Abstract
Additive manufacturing (AM) is a transformational digital manufacturing technology featured with rapidity, customizability, precision, and economy, which is fundamentally altering the way components are designed and manufactured. AM enables the freedom of design, and makes full use of complexity of geometry which “comes for free”. Applying AM technology to nuclear industry can yield advanced reactor designs with function and structure matched for the best thermal, hydraulic and mechanical performance. In this work, an AM-informed reactor core design with silicon carbide (SiC) matrix and tri-structural isotropic (TRISO) particle fuel is proposed and analyzed. The core is an integrated 3D-printing SiC bulk with helical cruciform coolant channels, and the UO2-bearing TRISO fuel particles are dispersed in the bulk. A multi-physics analysis framework for irregular geometry is developed to analyze and further optimize the reactor design. The TRISO particle positions are generated with discrete element method (DEM). The Reactor Monte Carlo code (RMC) and the commercial CFD software STAR-CCM+ are used for the neutronic and thermal-hydraulic analyses, respectively. RMC simulates the neutron transport to predict the effective multiplication factor and power distribution. STAR-CCM+ calculates the flow and heat transfer in coolant channels and heat conduction in solid matrix with the power distribution as the heat source. Preliminary results show that the power peaking factor FQ decreases below 1.65, the heat transfer area increases by 30.3% and the fuel peaking temperature decreases by 25 K. The optimized AM-informed design enjoys better neutronic and thermal-hydraulic performance than those with regular geometry.
Selection and Economic Benefit Analysis of Current Limiting Circuit Breaker in a Project of a Newly-Built Large Nuclear Chemical Plant
Authors: Gao Shuo , Tian Shizhong , Huang Zhi 
Published: November 2022

Abstract
With the continuous development of science and technology and the iteration of various types of low-voltage protection products, electrical designers are focusing not only on the reliability and security of the power supply system but also its economic benefits. From the calculation of electricity load in the early stage, to the selection of transformers, and then to the determination of the number of low-voltage distribution cabinets, as well as the selection of each feeder circuit protection devices, designers strive to come up with the most reliable and economical choice in each link through rigorous calculation and analysis. In the design and installation of large communities or factories, a small technical improvement can bring substantial economic benefits. Combined with engineering practice, this paper briefly introduces several protective appliances and their functions in low-voltage distribution circuits, the main differences between current limiting circuit breakers and ordinary circuit breakers are also revealed. The selection principle and basis of current limiting circuit breaker in the electrical design of a nuclear chemical engineering project are described in detail. The economic benefits and significance of selecting current limiting circuit breaker are compared and analyzed in this paper. The development prospects of current limiting circuit breaker selection in power supply and distribution system in large nuclear chemical plants and its key role for the whole plant in the future are discussed.
Implementation and Analysis of Thermal Hydraulics Feedback in High-Fidelity Neutron Transport Program HNET
Authors: Zhu Yanling , Hao Chen , Li Peijun , Zhou Xiaoyu 
Published: November 2022

Abstract
With the growing requirement of predicting reactor behavior in high-fidelity detail at practical conditions, it is urgent to accomplish thermal hydraulics (T-H) feedback in the high-fidelity neutron transport program HNET. For better convergence behaviors than Picard iteration, the Matrix Free Newton/Krylov (MFNK) method was employed to resolve neutronics and thermal-hydraulics coupling system. MFNK treats each subsystem as a black box within the Newton method framework, so it can facilitate the coupling procedure without surrendering efficiency or robustness. For the T-H feedback effects, a simplified internal thermal hydraulics model was adopted to provide T-H conditions for neutronics. The convergence behaviors of MFNK and Picard iteration were assessed through simple typical cases. Finally, the effectiveness of the coupling system was verified by the VERA problem #6. Numerical results demonstrate the efficiency and stability of MFNK compared with Picard iteration. Moreover, it turns out that the coupling system has a good performance in realistic cases.
Research and Treatment of Pressure Regulation Defects in Fuel Circuit of Emergency Diesel Engine of One Nuclear Power Plant
Authors: Feng Guo 
Published: November 2022

Abstract
For one nuclear power plant, the pressure of the diesel engine fuel oil circuit happened to fail to fit the design because of low fuel oil inlet pressure and fall-off of the O-ring of the pressure regulating valve. A lot of data has been collected from tests performed in different conditions during the commissioning phase of this type of diesel, and after a lot of work the wrong diameter of the orifice plate and the improper inner structure of the pressure regulating valve in the fuel oil returning circuit were proved to be the main reasons. This project tried to solve the problems in a systematic way. Single-variable method was used to find out that the inner hole diameter of the orifice plate was too large to maintain the fuel oil pressure at a high level even if the pressure regulating valve was blocked and the right diameter was found and tested on site. The inner part (valve disc) of the pressure regulating valve was found fail to keep the O-ring under the alterable pressure, then fail to regulate the fuel oil pressure. The structure of the valve disc was optimized to fit the alterable pressure working situation. Finally, the defect in fuel circuit of emergency diesel engine was solved by adjusting the bore diameter of the orifice plate and changing the structure of the pressure regulating valve disc. The relevant tests were successfully done after the modification implementation.
Shutdown Seal Improvement Feasibility Analysis for CPR1000 Reactor Coolant Pump
Authors: Hengrui Zhang , Qian Huang , Peiren Fu , Daoke Zhang , Rui Cai , Lei Sun , Qiuyang Chen , Feng Xue 
Published: November 2022

Abstract
After the Fukushima accident, there were two important feedbacks. One was the simultaneous occurrence of serious accidents of multiple units at the same plant site, and the other was the emergency diesel generator was unavailability under a serious accident. The events at Fukushima highlight the possibility that extreme natural phenomena could challenge the prevention, mitigation and emergency preparedness defense-in-depth layers. In the design of CPR1000 Nuclear power unit, the important contributor is the integrity of the reactor coolant system considering potential leaks from the reactor coolant pump seals following a station blackout (SBO) or loss of heat sink accident. After the SBO or loss of heat sink of whole plant occurs, normal main pump shaft seal injection will be lost. The hydrostatic test pump will start automatically to ensure emergency shaft seal injection. But the flow of hydrostatic test pump can only meet the requirements of one unit shaft seal injection, and the other unit shaft seal injection cannot be guaranteed, which may lead to loss of coolant accident (LOCA). If the LOCA is not controlled, it will lead to reactor core damage. At present, the improvement of shutdown seal used in the main pump of the new power plant can effectively solve this problem and reduce the frequency of reactor core damage. This paper mainly analyzes the technical characteristics of shutdown seal and the structure and function of main pump, so as to determine the feasibility and necessity of the improvement of shutdown seal of CPR1000 main pump.
State Monitoring of Electric Gate Valve of Nuclear Power Plant Based on Kernel Principal Component Analysis and Jensen-Shannon Divergence
Authors: Wang Xiaokun , Wang Hang , Deng Qiang , Xu Renyi 
Published: November 2022

Abstract
With the rapid development of the global nuclear power industry, the safety of nuclear power equipment has received increasing attention. In the event of an accident in its equipment, it is likely to cause significant losses and harm to personnel and the environment. Therefore, real-time monitoring of nuclear power plants has very important practical significance. Based on Kernel Principal Component Analysis (KPCA) and Jensen-Shannon (JS) divergence, equipment-level state monitoring of electric gate valves of nuclear power plants will be conducted. The traditional Principal Component Analysis (PCA) method, which can only reduce the dimensionality of nonlinear data, is improved by KPCA. The traditional statistics T2 and Q, which have the problem of low recognition rate of small faults and poor timeliness, can be greatly improved by the JS divergence. The main model process is to first perform KPCA dimensionality reduction on the data, calculate the JS divergence value of the data after dimensionality reduction, and compare the change trend of the JS divergence value to reflect the operating status of the electric gate valve, and perform real-time status monitoring. The experimental results show that the state detection method combining KPCA data dimensionality reduction and JS divergence has improved the accuracy of monitoring compared with other methods, and has better timeliness.
Preparation and Properties of Ceramic Solidified Product Containing Cs and Sr
Authors: Sang Hongji , Gu Zhengcheng , Cui Zheng , Zou Ruoxue , Wu Yan 
Published: November 2022

Abstract
137Cs and 90Sr in high level liquid waste, with high radioactivity, large heat-generating and relatively long half-life. In order to solve the problem of final disposal of 137Cs and 90Sr, natural mineral allophane was chosen as the base materials to synthesize the silicate ceramic solidified products with cold pressing/sintering method. The microstructure, phase composition and surface element distribution of the solidified products were analyzed. The solidification mechanism of the solidified products was also discussed. The surface of the solidified products appeared melting phenomenon after sintering, and the structure was more compact. CsAlSiO4, Sr2Al2SiO7 and SrAl2Si2O8, which can stably solidify Cs and Sr, were formed in the solidified products. The content of allophane in the solidified product has an important influence on the immobilization ratio of Cs and Sr. And at the sintering condition of 1 h duration time at 1200 °C, the immobilization ratios of Cs and Sr can reach 100 %. By increasing the content of cured substrate, the surface characteristics and crystallization properties of the solidified product can be improved, and the volume reduction effect was more obvious.