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2019 Transactions of the American Nuclear Society, ANS 2019

作者： Yuan, Peng Wu, Xiaoli Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610213 China

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The defect analysis technology study of reactor pressure vessel main bolt 27

The defect analysis technology study of reactor pressure ves...

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27th International Conference on Nuclear Engineering: Nuclear Power Saves the World!, ICONE 2019

作者： Zheng, Liangang Zhang, Rui Huang, Qian Bai, Xiaoming Yuan, Feng Lu, Xifeng Yuan, Yanli Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu Sichuan China

ISBN: (纸本)9784888983051

The existence of non conformances and defects are inescapable during the manufacture and in service of reactor pressure vessel (RPV). After being analyzed, most of the defects have no influence on the capability and safety of the structure if the analysis results satisfy criteria. The defects including main bolts defects, the non conformances of cladding thickness and structure dimension, structure scratch, material inclusion and so on. The typical and serious defects is main bolts defects, including thread defect, stress relaxation of bolt and the extreme case of bolt break. The results of stress, fatigue and seal performance of reactor pressure vessel will be changed considering these defects. Based on the mechanics theory and numerical simulation technique, a kind of stress, fatigue and seal performance analysis method about the defects effects is studied in this paper. The influence to the reactor pressure vessel induced by several kinds of bolt break accident is analyzed using this analysis method. Copyright © 2019 by JSME

关键词： Fatigue of materials

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Severe accident mitigation measure and severe accident sequence of SMR 27

Severe accident mitigation measure and severe accident seque...

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27th International Conference on Nuclear Engineering: Nuclear Power Saves the World!, ICONE 2019

作者： Yin, Shasha Yan, Sun Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu Sichuan China

ISBN: (纸本)9784888983051

Interest in evaluation of reactor response and off-site consequences following beyond design basis accidents has significantly increased after Fukushima. MELCOR is an advanced computational aid that has been widely used and adapted to various reactor designs for severe core damage analyses. This paper summarizes the core thermal hydraulic response for a hypothetical severe accident caused by station blackout at a small modular reactor using MELCOR. The results show that: After a severe accident in SMRs, the passive core cooling system can cool the core by the natural circulation of water, extend the time of core uncover. The time of bottom head failed delays nearly 64000s. So the passive core cooling system have great mitigation effect on the severe accident induced by station blackout. The analyses are focused on safety assessment of the reactor core for severe accidents and are part of the overall evaluation of safety features of the small modular reactor for residual risk posed by severe accidents. Copyright © 2018-2019 by JSME

关键词： Thermoelectric equipment

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Effect of Rolling Motion on Transient Behaviors of Condensation-Induced Water Hammer in an Equal-Height-Difference Passive Heat Removal system

SSRN

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SSRN 2022年

作者： Duan, Zhongdi Wang, Zhiwei He, Yanping Huang, Chao Li, Mingzhi Yuan, Yuchao Liu, Shiwen State Key Laboratory of Ocean Engineering Shanghai Jiao Tong University Shanghai200240 China School of Naval Architecture Ocean & Civil Engineering Shanghai Jiao Tong University Shanghai200240 China Institute of Marine Equipment Shanghai Jiao Tong University Shanghai200240 China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610213 China

The investigation of condensation-induced water hammer (CIWH) under ocean conditions is vital to the system safety of the offshore floating nuclear plants (OFNPs). This paper establishes a CIWH numerical model under rolling conditions based on the computational fluid dynamics (CFD) code combined with user-defined functions (UDF). First, the numerical model is validated with the experimental data. Then, the transient behaviors of the CIWH under various rolling parameters are numerically investigated. The results show that the rolling motion causes a significant steam slug and increases the steam-subcooled water contact area, which promotes the average condensation rate. The average condensation rate increases as the rolling angle increases and the rolling period decreases. Due to the larger average condensation rate and reverse flow induced by rolling motion, the pressure peak generated by the CIWH significantly increases compared to static conditions. Similarly, the pressure peak increases as the rolling angle increases and the rolling period decrease. Moreover, the pressure peak is proportional to the steam inlet velocity, and the location of the pressure peak gradually moves back to the outlet direction as the steam inlet velocity increases. These qualitative conclusions may serve as a reference for the CIWH simulations under ocean conditions. © 2022, The Authors. All rights reserved.

关键词： Condensation

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The sensitivity analysis of the core lower head molten pool model based on variance decomposition 27

The sensitivity analysis of the core lower head molten pool ...

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27th International Conference on Nuclear Engineering: Nuclear Power Saves the World!, ICONE 2019

作者： Li, Zhigang Liu, Wei Ming, Pingzhou Pan, Junjie An, Ping Liu, Dong Lu, Wei Yu, Hongxing Nuclear Power Institute of China Science and Technology on Reactor System Design Technology Laboratory ChengDu610041 China

ISBN: (纸本)9784888983051

The lower head molten pool model is an important model for the In Vessel Retention (IVR) effectiveness evaluation. It has the characteristics of complex relationship, many input parameters and large uncertainty. The traditional local sensitivity analysis method has the disadvantages of large amount of calculation and low efficiency when performing sensitivity analysis of complex models. Based on variance decomposition of the global sensitivity analysis method, the sensitivity analysis library(SALib) and IVR analysis code(CISER) developed by Nuclear Power Institute of China(NPIC) are adopt to analyze the sensitivity of the input parameters for four key parameters, such as the lower head wall heat flux ratio. In the paper, the sensitivity coefficient and influence tread of input parameters on key results are obtained, which provides a reference for the optimization of the lower head molten pool model and severe accident mitigation strategies. Copyright © 2018-2019 by JSME

关键词： Sensitivity analysis

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An automatic optimization method for nuclear piping layout and mechanical analysis 27

An automatic optimization method for nuclear piping layout a...

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27th International Conference on Nuclear Engineering: Nuclear Power Saves the World!, ICONE 2019

作者： Bai, Xiaoming Lu, Xifeng Zhang, Rui Shi, Kaikai Zheng, Liangang Huang, Qian Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu Sichuan China

ISBN: (纸本)9784888983051

The piping system in the nuclear power plant is extremely complex. The traditional trial method for the optimization is time-consuming and experience-dependent. Moreover, it rarely obtains the layout with best mechanical property using the trial method. In order to improve the traditional method, an automatic optimization method is proposed in present work. The genetic algorithm is employed during the mechanical analysis of piping system. In this method, both the piping layout and the piping support can be optimized concurrently. Moreover, based on the optimization method, a piping optimization theory assistant tool (POTATO) is developed. Some optimization instances in the nuclear power plant is shown in this paper. Comparing with the layout from the trial method, the one from present automatic method have more mechanical margin. The stress ratios at all the conditions are less than the limitation. Meanwhile, the time cost in present method is less than the trial method. Thus, present method is more efficient than the traditional method, and is significant for the piping layout in nuclear engineering. Copyright © 2018-2019 by JSME

关键词： Genetic algorithms

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A combined reactivity control pattern on a floating nuclear power plant 27

A combined reactivity control pattern on a floating nuclear ...

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27th International Conference on Nuclear Engineering: Nuclear Power Saves the World!, ICONE 2019

作者： Zhou, Bingyan Song, Danrong Chen, Zhang Chai, Xiaoming Xiang, Hongzhi Hu, Yuying Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu Sichuan China

ISBN: (纸本)9784888983051

In this study, a new combined reactivity control pattern is presented to lower the power peaking factor on a floating nuclear power plant. In the pattern the whole cycle length is divided into several boron-adjusting periods. During one boron-adjusting period, the soluble boron concentration remains unchanged and the core reactivity change is controlled by control rods. Boron concentration would only be changed at the end of a boron-adjusting period. In the whole cycle the boron concentration would be changed for two to four times, which is acceptable for a floating nuclear power plant. To get an optimized solution, a model of the combined reactivity control pattern was established, which contained boron concentration and control rod pattern (CRP) as coupled variables. A particle swarm optimization with novel mutation operator (PSONM) was applied to optimize the solution. The model was practiced based on a small modular reactor ACP100 core. The solution was optimized separately for equal division of two, three, four and five boron-adjusting periods. The results show that the combined reactivity control pattern can get a lower power peak factor than the original rod-controlled-only pattern without shortening the cycle length. Therefore, the safety of the floating nuclear power plant can be improved. Copyright © 2018-2019 by JSME

关键词： Nuclear power plants

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A calibration method for raman distributed temperature sensor in nuclear power plant

A calibration method for raman distributed temperature senso...

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2019 Transactions of the American Nuclear Society, ANS 2019

作者： Xiang, Meiqiong Liu, Yanyang Qing, Xianguo Zhu, Jialiang Wu, Qian Zhu, Biwei Xu, Sijie Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610213 China

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The Effect of Device Redundancy in Timeliness of Information

The Effect of Device Redundancy in Timeliness of Information

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IEEE International Conference on Communications (ICC)

作者： Kang Lang Zhengchuan Chen Nikolaos Pappas Howard H. Yang Yunjian Jia Tony Q. S. Quek School of Microelectronics and Communication Engineering Chongqing University Chongqing China State Key Laboratory of Integrated Services Networks Xidian University Xi'an China Department of Computer and Information Science Linköping University Linköping Sweden Zhejiang University/University of Illinois at Urbana-Champaign Institute Zhejiang University Haining China Information System Technology and Design Pillar Singapore University of Technology and Design Singapore

Emerging interaction-based Internet of Things (IoT) applications have stringent demand for timeliness, imposing critical challenges to the design of status update system. Using redundant devices to update the status of the same process is a promising way to improve timeliness, but this approach can result in out of order update arrivals, making it difficult to analyze timeliness. To that end, the present paper conducts a theoretical study toward the Age of Information (AoI) of a multi-queue status update system where multiple sensors observe one physical process and update a common monitor. Based on the stochastic hybrid systems method, the average AoI of the considered system is derived in closed form. The theoretical results are consistent with the simulation results, verifying the correctness of the theoretical analysis. It is shown that the logarithm of the average AoI is linearly decreasing with the logarithm of the number of sensors.

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Study on Acoustic-Induced Vibration (AIV) by reactor coolant pump at secondary support structure 48

Study on Acoustic-Induced Vibration (AIV) by reactor coolant...

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48th International Congress and Exhibition on Noise Control Engineering, INTER-NOISE 2019 MADRID

作者： Huang, Xuan Cai, Fengchun Wu, Wanjun Qi, Huanhuan Feng, Zhipeng Liu, Shuai Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu Sichuan China

ISBN: (纸本)9788487985317

The reactor coolant pump is the core equipment of nuclear power stations. During the operation of reactor coolant pump, acoustic pulsations are generated due to nonuniform flow within the reactor coolant pump and pass through the main pipe and coolant to the structure in the reactor. The acoustic pulsations can cause the vibration of the structure in the reactor, which is regarded as one of the main reasons of fatigue failure of the structure. In this paper, the finite element model of secondary support structure will be built and verified by test, then the pressure pulsations relation between the reactor coolant pump outlet and the secondary structure will be investigated, the pressure pulsation caused by the reactor coolant was calculated. Finally, the acoustic-induced vibration response of secondary support structure caused by the reactor coolant pump were obtained by harmonic analysis and high circle fatigue analysis will be done on the basis of the results. This study is helpful to understand the acoustic-induced vibration of the reactor coolant pump and to improve the hydraulic design of the reactor coolant pump. © INTER-NOISE 2019 MADRID - 48th International Congress and Exhibition on Noise Control Engineering. All Rights Reserved.

关键词： Harmonic analysis

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