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Development of high-strength,low-cost wrought Mg–2.0 mass% Zn alloy with high Mn content

Progress in Natural science:Materials International 2016年第6期26卷 630-635页

作者： Fusheng.Pan Jianjun Mao Gen Zhang Aitao Tang Jia She College of Materials Science and Engineering Chongqing University National Engineering Research Center for Magnesium Alloys Chongqing University Chongqing Research Center for Advanced Materials Chongqing Academy of Science and Technology Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China

Mg–Zn–Mn-based alloys have received considerable attention because of their high creep resistance, strength,and good corrosion resistance. The alloying element Mn in Mg–Zn-based alloys is commonly less than 1 wt%. In the present study, the effect of high Mn content(1 wt% and 2 wt%) on the microstructures and mechanical properties of Mg–2Zn–0.3Sr extruded alloy was investigated. The results revealed that the high Mn content significantly increased the ultimate tensile strength, tensile yield strength, compress yield strength, and yield asymmetry of the alloy without affecting its ductility. The dynamically recrystallized(DRXed) grains of Mg–2Zn–0.3Sr were remarkably refined because of the large amount of fine Mn precipitates in the homogenized alloy. The improved strengths were mainly attributed to the fine DRXed grains according to the Hall–Petch effect and to the large amount of spherical and < 0001 > Mn precipitates through the precipitation and dispersion strengthening. The fine DRXed grains and numerous Mn precipitates effectively suppressed the extension twining, substantially enhanced the compress yield strength, and resulted in improved anisotropy.

关键词： Magnesium alloy Mg–Zn alloy Microstructure Ultra-fine grains Second phase Mechanical properties

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Research on the domestic materials of forging Z12CN13 stainless steel for Hold-Down Spring of reactor

Research on the domestic materials of forging Z12CN13 stainl...

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Energy Materials 2014

作者： Li, Ning Ying, Luo Xin, Xia Hua, Du National Key Laboratory of Science and Technology on Reactor System Design Technology Chengdu610041 China

ISBN: (纸本)9781119027942

Effects of main alloying elements, quenching and tempering temperatures and quenching cooling velocity on mechanical properties of Z12CN13 stainless steel were obtained by controlling the material composition and optimizing the heat treatments. Evaluation of domestic forging Z12CN13 stainless steel for Hold-Down Spring of reactor had been completed, and all parameters met the requirement of design criteria national standard. The main alloying elements should be selected from upper limits to reach the technical standards of the strength and toughness properties. The strength, plasticity and toughness properties of domestic forging Z12CN13 stainless steel will decrease with increasing the quenching temperature. On the contrary, the strength decreases, plasticity and the toughness properties will increase with the increasing of tempering temperature. The tensile strength and impact toughness of materials are influenced obviously with the quenching cooling velocity. Improving the cooling velocity could meet the technical standards of the strength and toughness properties. The comprehensive mechanical properties and combination of strength and toughness is the best when the domestic forging Z12CN13 stainless steel is quenched at 985°C and then cooling fast and aged at 640°C.

关键词： Stainless steel

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Research of SSAE-GPC in Coordinated Control system of Nuclear Power Plant 5th

Research of SSAE-GPC in Coordinated Control System of Nuclea...

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5th International Symposium on Software Reliability, Industrial Safety, Cyber Security and Physical Protection of Nuclear Power Plant, ISNPP 2020

作者： Deng, Zhi-Guang Wu, Qian Zhu, Bi-Wei Lv, Xin Zhu, Jia-Liang Xu, Si-Jie Wang, Xue-Mei National Key Laboratory of Science and Technology on Reactor System Design Technology Chengdu610213 China

ISBN: (纸本)9789811634550

For the problem of poor control effect of conventional PID controllers in complex systems, this paper fully combines the advantages of deep learning in feature extraction, regression prediction, and predictive control in dealing with multivariable, strong coupling and other issues. Firstly, the predictive model controller is constructed through the stacked sparse auto-encoder (SSAE) deep network, and then the stepped generalized predictive control (SGPC) is used as the rolling optimization controller. Thus, the SSAE-SGPC neural network predictive controller is built. Finally, a simulation experiment is carried out in the dual-input dual-output multi-variable coordinated control system of nuclear power plant. Through a series of simulations such as set value disturbance, internal disturbance and external disturbance, compared with conventional PID control, SSAE-GPC controller has better control effect. © 2021, The Author(s), under exclusive license to Springer Nature Singapore Pte Ltd.

关键词： Controllers

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Research and design on the Control Rod Drive Mechanism Power Supply and Control Strategy for the Nuclear Power Plants Based on Silicon-Controlled Rectifier 10

Research and Design on the Control Rod Drive Mechanism Power...

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10th IEEE International Power Electronics and Motion Control Conference, IPEMC 2024 ECCE Asia

作者： Xu, A. Mingzhou Gao Zheng, B. Zhao, C. Zerun Ping Yang, D. Peng, E. Yusheng Science and Technology on Reactor System Design Technology Laboratory China Southwest Jiaotong University China

ISBN: (纸本)9798350351330

Due to the characteristics of large output current ripple and long current rise and fall times in the current rod control power supply basedsilicon-controlled rectifier, a decoupling nonlinear adaptive control is *** proposed control strategy can improve the rod control power supply of the speed of respond and the steady-state current accuracy. Practical application has shown that this method solves the output current ripple and the rise and fall time of the current, meeting the working requirements of rod control power supply in nuclear power plants. © 2024 IEEE.

关键词： Nuclear fuels

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Application of feedforward predictive control in DC furnace coordination system

Application of feedforward predictive control in DC furnace ...

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2020 International Symposium on Automation, Mechanical and design Engineering, SAMDE 2020

作者： Deng, Zhiguang Zhengxi, H.E. Biwei, Z.H.U. Jialiang, Z.H.U. Xin, L.V. Qian, W.U. National Key Labortory of Science and Technology on Reactor System Design Technology Chengdu610213 China

In order to solve the control difficulties such as large dynamic delay, serious coupling between machine and furnace and strong nonlinear, the feedforward and feedback predictive control is proposed to control the coordinated system on the basis of multi-model predictive control, and the static feedforward part designs the feed line feedforward of coal quantity against coal quality disturbance on the basis of the accurate balance of energy. Thus, the reference total fuel quantity under the target load instruction is determined, and the dynamic part fully considers the dynamic compensation of the unit energy storage, gives a certain overshoot fuel quantity to the boiler regulation, speeds up its response, and ensures the rapidity of the unit. The real-time simulation results of virtual DPU (Distributed Processing Unit) show that the designed feedforward-feedback predictive control can quickly respond to load change, improve the resistance to fuel disturbance, reduce the fluctuation of main steam pressure, and better ensure the safe, economic and stable operation of the coordination system. © Published under licence by IOP Publishing Ltd.

关键词： Feedback

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Research of Advanced Control Algorithm in Primary Loop Control system 5th

Research of Advanced Control Algorithm in Primary Loop Contr...

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5th International Symposium on Software Reliability, Industrial Safety, Cyber Security and Physical Protection of Nuclear Power Plant, ISNPP 2020

作者： Deng, Zhi-Guang Zhu, Bi-Wei Wu, Qian He, Peng Xiang, Mei-Qiong Xu, Tao Qing, Yue National Key Laboratory of Science and Technology on Reactor System Design Technology Chengdu610213 China

ISBN: (纸本)9789811634550

The primary loop system of nuclear power plant is a large-scale control system with serious coupling between loops and complex objects. In order to solve the problem of integrated control of primary loop core power, stabilizer pressure and water level, and steam generator water level, this paper applies improved predictive control algorithms to each sub-control system on the basis of mechanism modeling. For the strong non-linear problem of water level control, a multi-model predictive controller is constructed through linearization at each typical operating point. The simulation results show that the application of advanced predictive control algorithm to the primary loop control system has achieved good control performance, and it also lays a solid foundation for subsequent practical engineering applications. © 2021, The Author(s), under exclusive license to Springer Nature Singapore Pte Ltd.

关键词： Water levels

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Assessment of Coupled Effect of Steam Pipes on Seismic Analysis of reactor Coolant system 7

Assessment of Coupled Effect of Steam Pipes on Seismic Analy...

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7th International Conference on Environmental science and Civil Engineering, ESCE 2021

作者： Li, Lijuan Liu, Zhenyu Li, Hui Ye, Xianhui Ai, Honglei National Key Laboratory of Science and Technology on Reactor System Design Technology Chengdu Sichuan China

reactor coolant system is connected with auxiliary pipes, surge pipes and steam pipes. In order to facilitate the dynamic analysis of reactor coolant system, it is necessary to decouple the pipes connected with the reactor coolant system. Although some international standards stipulating requirements for the decoupling of subsystems from the main system, in order to further explore the influence of decoupling steam pipes, the seismic response of reactor coolant system with or without steam pipes is simulated by using ANSYS program. Firstly, the modal analysis of reactor coolant system is carried out to analysis the influence of steam pipes on the resonant frequency of reactor system. Then, the standardized seismic dynamic analysis of reactor system with or without steam pipes is carried out. The nozzle load (inlet and outlet nozzle load of main device, such as reactor pressure vessel, steam generator, pump) and support load (of main device such as reactor pressure vessel, steam generator, pump) are output and compared. According to nozzle load and support load, it can be seen that the steam pipes have little influence on the seismic analysis of reactor coolant system, i.e. steam pipes can be decoupled in the seismic analysis of reactor coolant system. © Published under licence by IOP Publishing Ltd.

关键词： Steam generators

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Transient fuel performance analysis of UO_(2)–BeO fuel with composite SiC coated with Cr cladding based on multiphysics method

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Nuclear science and Techniques 2023年第12期34卷 93-108页

作者： Chun‑Yu Yin Shi‑Xin Gao Sheng‑Yu Liu Rong Liu Guang‑Hui Su Li‑Bo Qian School of Nuclear Science and Technology Xi’an Jiaotong UniversityXi’an 710049China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of ChinaChengdu 610213China School of Electric Power South China University of TechnologyGuangzhou 510640China

The transient multiphysics models were updated in CAMPUS to evaluate the accident-tolerant fuel performance under accident *** is a fuel performance code developed based on *** simulated results of the UO_(2)–Zircaloy fuel performance under accident conditions were compared with those of the FRAPTRAN code and the experimental data to verify the correctness of the updated ***,multiphysics models of the UO_(2)–BeO fuel and composite SiC coated with Cr(SiC_(f)/SiC-Cr)cladding were implemented in ***,the fuel performance of the three types of fuel cladding systems under Loss of Coolant Accident(LOCA)and Reactivity Insertion Accident(RIA)conditions was evaluated and compared,including the temperature distribution,stress distribution,pressure evolution,and cladding failure *** results showed that the fuel temperature of the UO_(2) fuel under accident conditions without pre-irradiation was lower after being combined with SiC_(f)/SiC-Cr ***,the centerline and outer surface temperatures of the UO_(2)–BeO fuel combined with SiC_(f)/SiC-Cr cladding reduced further under accident *** cladding temperature increased after the combination with the SiC_(f)/SiC-Cr cladding under accident conditions with *** addition,the use of SiC_(f)/SiC-Cr cladding significantly reduced the cladding hoop strain and plenum pressure.

关键词： Accident condition Fuel performance UO2–BeO fuel SiCf/SiC-Cr cladding

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In-situ TEM study of the effect of hydrogen-helium ratio on defect characteristics in Fe9Cr1.5W0.4Si F/M steel during H_(2)^(+)&He^(+)simultaneous irradiation

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Journal of Materials science & technology 2024年第6期173卷 192-201页

作者： Dewang Cui Ziqi Cao Kun He Yuanming Li Xi Qiu Guang Ran College of Energy Xiamen UniversityXiamen 361102China Fujian Provincial Nuclear Energy Engineering Technology Research Center Xiamen 361102China Science and Technology on Reactor Fuel and Materials Laboratory Chengdu 610041China Nuclear Power Institute of China Chengdu 610041China Science and Technology on Reactor System Design Technology Laboratory Chengdu 610041China

The effect of hydrogen and helium interaction,especially H-He ratio,on the irradiation behavior of nuclear materials has not yet been ***,this is an important basis for evaluating the irradiation properties of nuclear materials and developing high irradiation resistant ***,30 keV H_(2)^(+)and He^(+)dual beams with four H-He ratios of 0:10,3:10,15:10,and 30:10 were used to irradiate the newly developed Fe9Cr1.5W0.4Si F/M steel in TEM to in-situ study the interaction and ratio effect of hydrogen and *** addition of H atoms significantly promoted the nucleation of dislocation loops and *** the early stage of irradiation,the average size and density of dislocation loops increased with the increase of H-He ***,the larger the H-He ratio,the easier it was to form a complex dislocation ***,the final saturation size of bubbles increased with the increase of H-He *** was first found that the swelling was affected by H concentrations,with high H concentrations slowing down the increase in *** a certain irradiation dose,a specific H-He ratio would lead to a swelling peak of Fe9Cr1.5W0.4Si F/M *** super-sized bubbles at grain boundaries(GBs)were found after H addition,resulting in a bigger swelling of GBs than the *** the swelling of the GBs and the matrix show a dependence on the H-He *** current work is of great significance for understanding the interaction between hydrogen and helium in nuclear materials.

关键词： Dose ratio effect Ferritic-martensitic steel Hydrogen&helium synergetic irradiation In-situ TEM

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Preliminary research on the irradiation-thermal-mechanical coupling behavior simulation method of FCM fuel

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International Journal of Advanced Nuclear reactor design and technology 2019年 1卷 51-56页

作者： Changbing, Tang Yongjun, Jiao Yuanming, Li Yi, Zhou Hua, Pang Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institution of China No. 328 Section 1 Changshun Avenue ChengduSichuan610041 China

FCM fuel (fully ceramic micro-encapsulated fuel) which is a promising ATF (accident tolerant fuel) candidate fuel. Establishing the irradiation-thermal-mechanical coupling behavior simulation method of FCM fuel accurately, would be very helpful for the optimal design of FCM fuel. In this research, TRSIO particles were assumed to be uniformly distributed in the silicon carbide (SiC) matrix, and the corresponding representative volume element (RVE) models were established. Based on ABAQUS commercial software, the thermal effects and irradiation effects of FCM fuel were all introduced into this numerical simulation by coding the UMAT subroutine, and then the irradiation - thermal - mechanical coupling behavior simulation method of FCM fuel was preliminary established. Compared with the simulation result of BISON software in open literature, the simulation results of this numerical simulation method are basically consistent with those of BISON software, the effectiveness of this numerical simulation method was preliminary verified. © 2019 Xi'an Jiaotong University

关键词： Irradiation

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