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

作者： Xie, Weibo Wu, Zhiqiang Zeng, Shan Li, Yutong Hu, Qingren Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu Sichuan China

ISBN: (纸本)9784888983051

The main trend in the development of the local power plant control system is to adopt digital safety DCS system with rather strong ability of self-diagnosis. Apart from the self-diagnosis, it is also necessary to carry out periodic test in certain frequency to check the potential fault within a system and to certify the completeness and effectiveness of reactor protection function. Regarding nowadays in-service or in-building power plant, there are some deficiency in the automation and simplication of the periodic test method, which still can be improved. Based on the research on the overall optimization of nuclear power plant safety DCS periodic test, an integrated and comprehensive periodic test method was put forward. This method, not only achieved a high degree of automation, supporting T2 online test, which increasing the system reliability, in order to save the cost of human and material during window phase in the overhaul life. Copyright © 2018-2019 by JSME

关键词： Nuclear power plants

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Effective Strengthening and Toughening in High Entropy-Alloy by Combining Extrusion Machining and Heat Treatment

SSRN

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SSRN 2022年

作者： Pu, Zhuo Cai, Songlin L. Dai, Lan-Hong Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610200 China State Key Laboratory of Nonlinear Mechanics Institute of Mechanics Chinese Academy of Sciences Beijing100190 China School of Engineering Science University of Chinese Academy of Sciences Beijing100049 China CAS Center for Excellence in Complex System Mechanics Beijing100190 China China Electric Power Research Institute State Grid Corporation of China Beijing100192 China

This study presents an effective strengthening and toughening approach to improve tensile properties of CrMnFeCoNi high entropy-alloy (HEA) by combining extrusion machining and short-time heat treatment. After such process, preliminary structural heterogeneity has been achieved in this alloy, of which the strength is significantly enhanced while good elongation is retained compared to its homogeneous counterparts. Microstructural characterization reveals varying degrees of partial recrystallization with more or less dislocation density for HEAs under different treatment. It is such structural inhomogeneity that imparts this alloy with good strain hardening ability at high flow stress levels. © 2022, The Authors. All rights reserved.

关键词： Strain hardening

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Numerical methods research and fracture failure analysis for critical crack size 27

Numerical methods research and fracture failure analysis for...

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

作者： Shi, Kaikai Yuan, Yanli Chen, Jianguo Bai, Xiaoming Zheng, Liangang Lu, Xifeng Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu Sichuan China

ISBN: (纸本)9784888983051

Safety analysis of fracture mechanics is an important part of structural integrity assessment and critical crack size is one ofmain concerns for fracture failure. From the perspective of mathematics, the calculation of critical crack size is the solving and analysis for nonlinear equation f(x)=0. About the solution of nonlinear equation, there are the bisection method, the false position method, the simple iteration method, the newton method and the secant method. For the complexity of load imposed on nuclear power equipment, there may be the multi-value phenomenon of critical crack size in analyzing fracture failure problems. Considering that the corresponding mechanical states of the shallow crack and the deep crack are different when the fracture failure occurs, the crack toughness factor needs to be considered in the fracture failure analysis of shallow cracks. Connecting the bottom head of nuclear power plant main equipment, the pipeline large break transient of accident condition and the welding research council bulletin 175, the algorithm and the convergence of multi-value phenomenon of critical size are studied. Using the crack toughness factor, the fracture failure critical crack size is deeply analyzed and discussed. The algorithm and fracture failure theory in the present study can provide the technical support for the prolonging operation life projects of nuclear power plant main equipment in future. Copyright © 2018-2019 by JSME

关键词： Numerical methods

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Analysis of PWR SBO scenarios under refueling condition using MAAP5

Analysis of PWR SBO scenarios under refueling condition usin...

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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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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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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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Acoustic performance of fluid-filled pipe with periodic helmholtz resonator considering lateral wall elasticity

Acoustic performance of fluid-filled pipe with periodic helm...

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2019 Transactions of the American Nuclear Society, ANS 2019

作者： Zeng, Qingna Wang, Donghui Zang, Fenggang Zhang, Yixiong Science and Technology on Reactor system Design Technology Laboratory Nuclear Power Institute of China Chengdu610213 China

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Improvement and optimization of naspic SVDU engineering design in nuclear power plant according to NUREG0700 27

Improvement and optimization of naspic SVDU engineering desi...

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

作者： Li, Yutong Xie, Weibo Ma, Quan Zhang, Yu Zhao, Yang Hu, Qingren Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu Sichuan China

ISBN: (纸本)9784888983051

Safety Video Display Unit (SVDU) provides the interface between operator and reactor protection system (RPS).SVDU not only displays the signals and parameters which related to safety function but also support the operator to send control commands to *** a safety classified equipment and part of human machine interface (HMI), engineering design of SVDU should be addressed in plant safety with regarding to the HMI related standards. NUREG0700 is the guideline of human-system interface design, using which as a tool, a lot of improvement has been made to optimize the design of NASPIC SVDU: Display view arrangement, PAMS parameter curves, alarming etc. were improved and optimized. Then conformity analysis of NASPIC SVDU regarding NUREG0700 has been made, and equipment qualification was conducted. Finally NASPIC SVDU was prone to be in conformity with NUREG0700 and was able to achieve effective human-computer interaction. Copyright © 2018-2019 by JSME

关键词： Nuclear fuels

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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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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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