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

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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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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Openfoam Simulation of Damping Controlled Fluidelastic Instability 23rd

Openfoam Simulation of Damping Controlled Fluidelastic Insta...

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23rd Pacific Basin Nuclear Conference, PBNC 2022

作者： Feng, Zhipeng Liu, Shuai Qi, Huanhuan Huang, Xuan Lv, Xi Science and Technology On Reactor System Design Technology Laboratory Nuclear Power Institute of China Sichuan Chengdu China

ISBN: (纸本)9789811987793

A parallel triangular tube bundle with a single flexible tube is taken as the object, and the most dangerous mechanism of flow induced vibration in the tube bundle structure, fluidelastic instability (FEI) is studied based on the open-source CFD tool OpenFOAM. Firstly, a numerical model describing the fluid-structure interaction of the tube bundle is established by combining computational fluid dynamics with the vibration equation of the tube;Then, the variation of lift coefficient, drag coefficient, lateral displacement and lift spectrum with velocity is discussed. The results show that the open-source tool OpenFOAM can be easily used to predict the fluidelastic instability of tube bundle. © 2023, The Author(s), under exclusive license to Springer Nature Singapore Pte Ltd.

关键词： Computational fluid dynamics

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An Unsupervised Learning-Based Framework for Effective Representation Extraction of reactor Accidents 23rd

An Unsupervised Learning-Based Framework for Effective Repre...

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23rd Pacific Basin Nuclear Conference, PBNC 2022

作者： Li, Chengyuan Li, Meifu Qiu, Zhifang Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Sichuan Chengdu China

ISBN: (纸本)9789811987793

Compared to knowledge-based diagnostic systems, data-based methods tend to perform better in terms of speed and accuracy in diagnosing reactor accidents, and have significant advantages in terms of scalability of models. With the increasing use of high-precision system analysis programs in nuclear engineering, the number of high-fidelity computational data for accident simulation is exploding. Therefore, an algorithm that can achieve both automatic extraction of low-dimensional features from the data and guarantee the validity of the features is needed to improve the performance and confidence of the accident diagnosis system. This study proposes an autoencoder-based autonomous learning framework, namely Padded Auto-Encoder (PAE), which is able to automatically encode accident monitoring data that has been noise-added and with partially missing data into low-dimensional feature vectors via a Vision Transformer-based encoder, and to decode the feature vectors into noise-free and complete reconstructed monitoring data. Thus, the encoder part of the framework is able to automatically infer valid representations from partially missing and noisy monitoring data that reflect the complete and noise-free original data, and the representation vectors can be used for downstream tasks for accident diagnosis or else. In this paper, LOCA of HPR1000 was used as the study object, and the PAE was trained by an unsupervised method using cases with different break locations and sizes as the dataset. The encoder part of the pre-trained PAE was subsequently used as the feature extractor for the monitoring data, and several basic statistical learning algorithms for predicting the break locations and sizes. The results of the study show that the pre-trained diagnostic model with two stages has a better performance in break location and size diagnostic capability with an improvement of 41.62% and 80.86% in the metrics respectively, compared to the diagnostic model with end-to-end model st

关键词： Location

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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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Research on Output Torque Optimization of Ultrasonic Motor Based on Genetic Algorithm 17th

Research on Output Torque Optimization of Ultrasonic Motor B...

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17th Annual Conference of China Electrotechnical Society, CES 2022

作者： Li, Xiaoyang Tang, Jiankai Yang, Bo Wu, Hao Xu, Qiwei Yang, Yun The State Key Laboratory of Power Transmission Equipment & System Security and New Technology Chongqing University Chongqing400030 China Science and Technology on Reactor System Design Technology Laboratory Chengdu610041 China

ISBN: (纸本)9789819904501

Aiming at the problem of small output torque in the design process of traveling to wave ultrasonic motor, this paper proposes a method of optimizing the output torque of traveling to wave ultrasonic motor based on multi-objective genetic algorithm. It can effectively improve the output torque of the motor. Firstly, the structure and operation mechanism of the motor is introduced, and the key factors affecting the output torque of the motor is analyzed reasonably;then the parameterized model of the stator of the ultrasonic motor is established by WORKBENCH, the sample points are randomly selected in the design variable space, and the combination of each sample point is obtained. The modal analysis and harmonic response analysis of the stator structure was carried out to establish the response surfaces model of the stator;finally, the global optimal solution was obtained based on the response surface model by using the multi-objective genetic algorithm (MOGA). The results show that the output torque of the ultra-sonic motor can be increased by 8% before and after optimization. © 2023, Beijing Paike Culture Commu. Co., Ltd.

关键词： Stators

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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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Investigation on crack propagation of SiCf/SiC composite via PF-CZM method

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International Journal of Advanced Nuclear reactor design and technology 2024年第1期6卷 9-13页

作者： Qiu, Bowen Li, Zhuang Liang, Shuang Zhang, Cheng Wei, Chong Xiao, Zhong Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610213 China School of Mechanical Civil Engineering and Architecture Northwestern Polytechnical University Xi'an710072 China Science and Technology on Thermo Structural Composite Materials Laboratory Northwestern Polytechnical University Xi'an710072 China

SiCf/SiC composite materials had become a research hotspot in the fields of aerospace materials and nuclear materials in recent years. This materials have high strength, radiation resistance, high temperature resistance and oxidation resistance. However, SiC is a typical brittle material prone to being damaged due to stress concentration in some cases, leading to cracks and even component failure. Therefore, this study employed an accurate and efficient numerical simulation method, phase field cohesive zone method (PFCZM), to investigate the microscopic damage behavior of SiCf/SiC composite materials. This study aimed to illustrate the influence of the thickness of pyrolytic carbon (PyC) interface layer on the failure behavior of SiCf/SiC composite materials. Firstly, this study conducted relevant verification of PFCZM. Secondly, numerical simulations of crack initiation, crack propagation, and final damage were performed by establishing a representative volume element (RVE). Observing the simulation results, it was noted that at low strain, when PyC was thicker, the SiCf/SiC composite materials exhibited slight damage to both the SiC matrix and PyC interface. Conversely, when PyC was thinner, the SiCf/SiC composite materials displayed only slight damage to PyC interface. As the PyC thickness increases, the damage to PyC interface decreases, while the damage to SiC becomes more prominent. Simultaneously, a competitive relationship in the damage behavior of SiCf/SiC composite materials can be found after the SiC matrix has cracked: When PyC is thinner, the damage primarily manifests as interlayer delamination between the SiC matrix and the SiC fiber. Conversely, when PyC is thicker, it manifests in PyC failure. Finally, this study concluded that, in this investigation, a PyC interface layer with a thickness of 0.1 μm exhibited the most effective protective effect on SiCf/SiC composite material. © 2024 Xi'an Jiaotong University

关键词： Crack propagation

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Study on the Influence of reactor system LOCA Modeling Mode on Dynamic Analysis 23rd

Study on the Influence of Reactor System LOCA Modeling Mode ...

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23rd Pacific Basin Nuclear Conference, PBNC 2022

作者： Shuai, Liu Xuan, Huang Zhipeng, Feng Zhongxiu, Zeng Huanhuan, Qi Xiaozhou, Jiang Harbin Engineering University Harbin China Science and Technology on Reactor System Design Technology Laboratory Sichuan Chengdu China

ISBN: (纸本)9789811987793

The reactor system has a large scale of components. And the system dynamic analysis model for loss of coolant accidents (LOCA) contains many non-linear factors. The transient calculation analysis takes a long time and the convergence is difficult to guarantee. In this paper, two different analysis models (complete three-loop model and single-loop model) of the first loop system of reactor are established for comparative analysis. The analysis shows that the calculation error of the main position load in the two different model is within 10%. The decoupling influence of the wave tube and the reactor inlet and outlet on the system dynamic characteristics is within an acceptable range. When the single loop model is used, it is necessary to decouple the reactor pressure vessel and pressurizer. And the simplified boundary is adapted to simulate its effect on the whole loop. The single-loop dynamic analysis model based on the decoupling criterion can greatly reduce the calculation scale and meet the calculation accuracy, which can be used to calculate the LOCA dynamic characteristics of the reactor system quickly. The model can also be used for seismic analysis of reactor system. And using this model for reactor system dynamic analysis can greatly improve the calculation efficiency. Please copy you summary here. © 2023, The Author(s), under exclusive license to Springer Nature Singapore Pte Ltd.

关键词： Loss of coolant accidents

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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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