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30th International Conference on Nuclear Engineering, ICONE 2023

作者： Zhu, Ye Liang, Qiao Xianwei, Liao Zhiyun, Cai Minghao, Liu Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Sichuan Chengdu China

ISBN: (纸本)9784888982566

The once through steam generator has great heat transfer capacity and compact structure. Some small module reactors (SMR) applied once through steam generatorfor its high thermal efficiency. The thermal hydraulic characteristic of once through steam generator is different from traditional natural cycle steam generator as it doesn't contain water level. Thus the related feed water system is important for the steady operation of the power plant. Base on the thermal hydraulic characteristic of once through steam generator, SMR feed water system that contains process and control model is developed using APROS. The operating parameters are simulated under both different steady states and transient condition. The results show that both the thermal hydraulic and automatic control parameters of SMR feed water system can be simulated accurately in APROS. Reference and guidance can be provided for system optimization and safety analysis. © 2023 by JSME.

关键词： Water levels

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Response Time Analysis Method of Safety-Level DCS in Nuclear Power Plant Based on Probability Theory 1

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7th International Symposium on Digital Instrumentation and Control technology for Nuclear Power Plant, SICPNPP 2023

作者： Wen, Jing He, Xian-Jian Chen, Zhao Wan, Zi-Yuan Wu, Zhi-Qiang Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610000 China

ISBN: (数字)9789819934553

ISBN: (纸本)9789819934546

Aiming at the incompleteness and inadequacy of the response time analysis method of the current nuclear power plant safety-level digital-control system (DCS), this paper proposes an analysis method based on probability theory for the response time of nuclear power plant safety-level DCS. Firstly, according to the structure and characteristics of the safety-level DCS, the probability model of the response time of the safety-level DCS is established, and the calculation method of the probability distribution of the response time is obtained. Then, according to the obtained calculation method and the parameters of the engineering example, the rang of the response time and the probability density function are calculated, and the response time of the engineering example is simulated, and the simulated response time rang, frequency and probability density function fitting curve are obtained. Finally, the response time of the engineering example is tested by the response time automatic test device, a large amount of data with statistical significance is obtained. After the data is processed, the range, frequency and probability density fitting curve of the response time are obtained. By comparing the theoretical calculation results, simulation results and test results of engineering examples. The correctness and feasibility of the analysis method are verified. © 2023, The Author(s), under exclusive license to Springer Nature Singapore Pte Ltd.

关键词： Nuclear power plants

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Effect of Corrosion Damage on Structural Failure Models Under Different Boundry Conditions 23rd

Effect of Corrosion Damage on Structural Failure Models Unde...

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23rd Pacific Basin Nuclear Conference, PBNC 2022

作者： Qian, Sheng Sun, Yu Huangfu, Yuzhao Zhang, Ke Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Sichuan Chengdu China

ISBN: (纸本)9789811987793

Nuclear energy is an important means to deal with the energy crisis and global climate change, and the long-term safe operation of nuclear power plants strongly depends on the service behavior of structural materials. Nuclear power equipment is subject to complex multiple loads in actual situations, especially under compression, and even the buckling of tiny structures may have important safety hazards. Experience has shown that corrosion is an unavoidable result of nuclear power structures, and structures damaged by corrosion will cause internal stress redistribution under compression, resulting in changes in failure modes. This paper takes the most common uniform corrosion and pitting corrosion as the research objects, and explores the effects of different corrosion forms and corrosion parameters on the buckling failure of compression members under the same corrosion rate. © 2023, The Author(s), under exclusive license to Springer Nature Singapore Pte Ltd.

关键词： Nuclear fuels

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Semi-empirical and semi-quantitative lightweight shielding design method

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Nuclear science and Techniques 2023年第3期34卷 109-124页

作者： Song-Chuan Zheng Qing-Quan Pan Huan-Wen Lv Song-Qian Tang Xiao-Jing Liu College of Smart Energy Shanghai Jiao Tong UniversityShanghai 200240China School of Nuclear Science and Engineering Shanghai Jiao Tong UniversityShanghai 200240China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of ChinaChengdu 610200China

The lightweight shielding design of small reactors is a popular research *** on a small helium-xenon-cooled solid reactor,the effects of neutron and photon shielding sequence and the number of shielding layers on the radiation dose were first *** was found that when photons were shielded first and the number of shielding layers was odd,the radiation dose could be significantly *** reduce the weight of the shielding body,the relative thickness of the shielding layers was optimized using the genetic *** optimized scheme can reduce the radiation dose by up to 57%and reduce the weight by 11.84%.To determine the total thickness of the shielding layers and avoid the local optimal solution of the genetic algorithm,a series of formulas that describes the relationship between the total thickness and the radiation dose was developed through large-scale calculations.A semi-empirical and semi-quantitative lightweight shielding design method is proposed to integrate the above shielding optimization method that verified by the Monte Carlo ***,a code,SDIC1.0,was developed to achieve the optimized lightweight shielding design for small *** was verified that the difference between the SDIC1.0 and the RMC code is approximately 10%and that the computation time is shortened by 6.3 times.

关键词： Small reactor Lightweight Shielding calculation Genetic algorithm

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Analytical Method Research of Source Term for Steam Generator Tube Rupture Accident of Small Modular reactor 23rd

Analytical Method Research of Source Term for Steam Generato...

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23rd Pacific Basin Nuclear Conference, PBNC 2022

作者： Xia, Ming-ming Wang, Jun-long Zhu, Jian-ping Tian, Chao Liu, Jia-jia Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Sicuan Chengdu China

ISBN: (纸本)9789819910229

Based on the full reference to the existing engineering practice and safety review experience, and considering the actual design characteristics of the small modular reactor ACP100, a set of source term analysis method suitable for ACP100 of the steam generator tube rupture (SGTR) accident is proposed, and the source term analysis and consequence evaluation of ACP100 SGTR accident was carried out using this method. The analysis shows that the radiological consequences of the accident source term calculated by this method meet the acceptance criteria of small modular reactor. The analysis results of this article can provide support for the follow-up review of accident source term of ACP100. © 2023, The Author(s).

关键词： Accidents

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Research on Thermal-Hydraulic Model and Characteristics of Lead Cooled Traveling Wave reactor 23rd

Research on Thermal-Hydraulic Model and Characteristics of L...

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23rd Pacific Basin Nuclear Conference, PBNC 2022

作者： Zhang, Yaoxiang Zhang, Xue Yu, Hongxing Du, Sijia Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610213 China

ISBN: (纸本)9789819910229

Traveling wave reactor (TWR) is an advanced nuclear power system, which can keep the total amount of fissionable nuclides constant during its lifetime through the transformation of fissionable nuclides in the reactor, so as to maintain the stability of the effective proliferation factor of the reactor core. As a coolant, lead has low neutron moderation ability and capture cross section. At the same time, it also has excellent thermal characteristics. Lead cooled traveling wave reactor is a reactor with high inherent safety and good application prospects. The research on reactor core mainly focuses on the physical calculation of the reactor, and the analysis of its thermal hydraulic model and characteristics is relatively scarce, which limits the design and safety analysis of the advanced reactor. Therefore, it is necessary to investigate the thermal hydraulic model of lead cooled traveling wave reactor. In this paper, a thermal hydraulic model suitable for the core of lead cooled traveling wave reactor is proposed. Considering the axial particularity of heat release from the core of traveling wave reactor, the model considers the axial thermal conductivity of fuel rods ignored by other thermal hydraulic models. Due to the strong thermal conductivity of core coolant, the thermal model also considers the heat transfer of fuel assemblies and the flow distribution of core flow. The research shows that the thermal hydraulic model is suitable for the analysis of thermal hydraulic characteristics of lead cooled traveling wave reactor, the thermal hydraulic characteristics are studied by using the thermal hydraulic model. © 2023, The Author(s).

关键词： Thermal conductivity

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Influence of Impedance Characteristics of Perforated Plate on Pump-Induced Pulsation in reactor 23rd

Influence of Impedance Characteristics of Perforated Plate o...

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23rd Pacific Basin Nuclear Conference, PBNC 2022

作者： Yu, Sun Fengchun, Cai Sheng, Qian Yuzhao, Huangfu Ke, Zhang Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Sichuan Chengdu China

ISBN: (纸本)9789811988981

In primary circuit, pump-induced pulsation (PIP) of coolant may cause fatigue failure of reactor components. In this article, a three-dimensional numerical model of PIP was established by COMSOL, theory of acoustic impedance of perforated plate was introduced and the perforated plate construction in reactor, such as flow baffle plate (FBP) was modeled by impedance boundary condition. In this paper, the calculation results of PIP pressure load were compared considering and without considering the impedance boundary condition. This study finds that differences of pressure amplitude on both sides of impedance boundary are significant when excitation frequency is coupled with the acoustic mode of computational model or when excitation frequency is relatively high. In addition, there are great differences of acoustic resonance frequency between two calculation models. © 2023, The Author(s), under exclusive license to Springer Nature Singapore Pte Ltd.

关键词： Boundary conditions

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Effect of He/dpa ratio on bubble characteristics in Fe9Cr1.5W0.4Si F/M steel during irradiation and annealing

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Journal of Materials science & technology 2024年第30期197卷 17-24页

作者： Dewang Cui Kefei Pei Ziqi Cao Yipeng Li Yifan Ding Yuanming Li Shichao Liu Guang Ran College of Energy Xiamen UniversityXiamen361102China Science and Technology on Reactor System Design Technology Laboratory Chengdu610213China College of Chemistry and Chemical Engineering Xiamen UniversityXiamen361005China Fujian Provincial Nuclear Energy Engineering Technology Research Center Xiamen361102China Nuclear Power Institute of China Chengdu610213China

The He production rate(i.e.,He/dpa)in nuclear reactors strongly affects the degradation of material *** is an important but not yet fully understood ***,the effect of He/dpa on bubble char-acteristics in Fe9Cr1.5W0.4Si ferrite-martensitic(F/M)steel was in situ studied during 400 keV Fe^(+)and 30 keV He^(+)dual-beam irradiation at 723 K with three ratios of 100,500,and 2500 appm He/dpa and subsequent stepwise annealing using transmission electron microscopy(TEM).He/dpa strongly affected the bubble *** irradiation,the higher the He/dpa,the smaller the size of irradiated bubbles,but the higher their ***,He/dpa didn’t affect the final saturation size of irradiated bubbles for all three cases,which was-2.2 *** annealing,high He/dpa caused large,immobile,dense polyhedral bubbles with a wider bubble size distribution,while low He/dpa caused small,low-mobility,and relatively low-density spherical *** was found that the higher the He/dpa ratio,the greater the swelling during irradiation and annealing,and annealing further enhanced the ***,the tunnel structure was first found in body-centered cubic(BCC)F/M steel during in-situ *** current work provides valuable and potential insights for further understanding the He/dpa effects in materials serving in different nuclear reactors.

关键词： He effect Ferritic-martensitic steel He bubble Ion irradiation Annealing

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REVIEW OF RISK-INFORMED SAFETY design CONCEPT, METHOD AND APPLICATION 30

REVIEW OF RISK-INFORMED SAFETY DESIGN CONCEPT, METHOD AND AP...

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30th International Conference on Nuclear Engineering, ICONE 2023

作者： Gong, Lanxin Zhong, Mingjun Peng, Changhong School of Nuclear Science and Technology University of Science and Technology of China Hefei China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China

ISBN: (纸本)9784888982566

Since the NRC issued the PSA policy in 1995, PSA containing risk information has been widely used in supervision and design. Subsequently, the concept of "risk-informed" has gained more and more recognition, and its application has expanded from the initial equipment maintenance to the current design. By analyzing the regulations, standards and other guidance documents of various countries and institutions on safety design, we reviewed the history of risk-informed safety design concepts, methods and their applications, especially the progress of LMP projects in the United States in recent years. We put forward prospects for the development and application of risk-informed concepts and methods in new reactors in the future. © 2023 by JSME.

关键词： design

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Study on Multi-Motor Synchronization Control for Nuclear Island Main Equipment Transporter Based on Improved Virtual Line-Shafting Structure 11th

Study on Multi-Motor Synchronization Control for Nuclear Isl...

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11th Frontier Academic Forum of Electrical Engineering, FAFEE 2024

作者： Sheng, Beining Song, Beibei An, Yanbo Du, Hua Cui, Shumei School of Electrical Engineering and Automation Harbin Institute of Technology Harbin150001 China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610213 China

ISBN: (纸本)9789819788231

The existing drive method of the main equipment transporter of the nuclear island makes the transporter have frequent jamming problems when the transporter going straight and turning, which affects the safety during the transporter of the main equipment. Therefore, it is necessary to design an on-board multi-motor synchronous system with differential synchronization function. Although the traditional virtual line-shafting structure has the ability of same-speed synchronization and differential synchronization, there is a large synchronization error. For this reason, this paper proposes a multi-motor synchronous control method based on the improved virtual line-shafting structure, which can effectively solve the inherent synchronous delay problem of the virtual line-shafting. First, analyze and research the combination method of virtual line-shafting structure and deviation coupling structure, and designs the electronic differential to be able to adapt to the requirements of same-speed and differential synchronization. Then the influence of virtual line-shafting parameters on the system is analyzed using the root locus method, and a simplified parameter model of the virtual line-shafting is established, on the basis of which an improved virtual line-shafting control method with low synchronous delay is proposed. The results show that this method can significantly reduce the synchronization error compared with the traditional virtual line-shafting. © Beijing Paike Culture Commu. Co., Ltd. 2025.

关键词： Root loci

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