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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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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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Research on boundary disposal of pipes decoupling based on PEPs 27

Research on boundary disposal of pipes decoupling based on P...

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

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

ISBN: (纸本)9784888983051

Generally, one of the remarkable characteristics is: compared with the main pipe, the sizes of the auxiliary piping are smaller, which makes the auxiliary piping satisfy the criteria to be decoupled from the main piping. So when the auxiliary pipes connected to the main piping are analyzed, there are two methods: one is that the auxiliary pipe is analyzed connected to the main piping, and the other is that the auxiliary pipe is analyzed not to be connected to the main pipe. Because the response spectrum at the location of the connection of the auxiliary pipes and the main piping is lacked, the previous choice is usually adopted. At present, the main software used for analyzing the stress of the pipes in nuclear power station is PEPS, but when the PEPS is adopted to analyze the small pipes, if the small pipes are analyzed connected to the main pipe, the boundaries of the thermal analysis and those of the seismic analysis are different, so the two kinds of analysis cannot be carried out at the only one model. That is, when the small pipe is analyzed connected to the main piping, two models are needed for thermal analysis and the seismic analysis, and then the results of the thermal analysis and the seismic analysis are distilled separately, and the results are combined according to determinate mode, and then the combined results are evaluated according to the criterion. Consequently, not only is the model complex, but also the combination of the seismic stress and the thermal stress is not convenient in the previous method. In the latter method, the auxiliary pipe is not connected to the main piping, accordingly, not only is the modal simple, but also the combination of the seismic stress and the thermal stress is convenient. The only problem which is needed to be solved is the seismic response spectrum and the displacements, and an approach to solve the problem is presented in the article. A response spectrum calculation method is obtained based on the single oscilla

关键词： Thermoanalysis

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Preliminary evaluation of fecral-UN fuel rod performance 14

Preliminary evaluation of fecral-UN fuel rod performance

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14th International Nuclear Fuel Cycle Conference, GLOBAL 2019 and Light Water reactor Fuel Performance Conference, TOP FUEL 2019

作者： Gao, Shixin Duan, Zhengang Zhou, Yi Chen, Ping He, Liang Liu, Shichao Zhang, Kun Qiu, Xi Zhang, Ruiqian Science and Technology on Reactor System Design Technology Laboratory No.328 Section 1 Changshun Road Shuangliu County Chengdu610213 China Nuclear Power Institute of China No.328 Section 1 Changshun Road Shuangliu County Chengdu610213 China

After the accident at the Fukushima Daiichi nuclear power plant, many institutions around the world developed the development of accident-tolerant fuel (ATF). Many institutions in China have been conducting research and development on accident tolerant fuels. NPIC plays the most important role in the CNNC Accident Tolerant Fuel (ATF) program, which aims to develop advanced fuels and claddings that have improved performance under normal conditions, design basis accidents (DBA) and beyond design basis accidents (BDBA). Due to its excellent corrosion resistance, FeCrAl stainless steel has become one of the most important ATF cladding concepts. The FeCrAl concept provides the operator with some extra coping time to deal with accident conditions compared to traditional zirconium-based cladding. Uranium mononitride has been identified as an alternative fuel with higher uranium density and thermal conductivity when compared to UO2. In NPIC ATF campaign, FeCrAl is one of the advanced cladding concepts and UN is one of the advanced fuel candidates. NPIC has developed about 4m full-length thin-walled FeCrAl tubes with several varied chemical composition as fuel rod cladding. NPIC also prepared UN fuel pellets by sintering. This paper focuses on in-pile performance analysis of UN fuel rod with FeCrAl cladding, so as to make suggestions for future FeCrAl-UN fuel rod research, preparation and application. This paper uses FUPAC code to make analysis on fuel rod in-pile performance. The simulation uses the most recently known material properties. Some analyses of FeCrAl-UN fuel rod performance at steady state conditions are performed. And the results show that FeCrAl-UN fuel rod has good performance in aspects such as fuel temperature, fission gas release, fuel rod internal pressure, etc. Finally, future research and development work are discussed. Copyright © GLOBAL 2019 - International Nuclear Fuel Cycle Conference and TOP FUEL 2019 - Light Water reactor Fuel Performance Confere

关键词： Ternary alloys

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Active learning for the power factor prediction in diamond-like thermoelectric materials

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npj Computational Materials 2020年第1期6卷 254-260页

作者： Sheng Yasong Wu Jiong Yang Wencong Lu Pierre Villars Wenqing Zhang Materials Genome Institute Shanghai University200444 ShanghaiChina Qianweichang College Shanghai University200444 ShanghaiChina Department of Chemistry College of ScienceShanghai University200444 ShanghaiChina Material Phases Data System CH-6354 VitznauSwitzerland Department of Physics and Shenzhen Institute for Quantum Science&Technology Southern University of Science and Technology518055 ShenzhenGuangdongChina Guangdong Provincial Key Lab for Computational Science and Material Design and Shenzhen Municipal Key Lab for Advanced Quantum Material and DeviceSouthern University of Science and Technology518055 ShenzhenGuangdongChina

The Materials Genome Initiative requires the crossing of material calculations,machine learning,and experiments to accelerate the material development *** recent years,data-based methods have been applied to the thermoelectric field,mostly on the transport *** this work,we combined data-driven machine learning and first-principles automated calculations into an active learning loop,in order to predict the p-type power factors(PFs)of diamond-like pnictides and *** active learning loop contains two procedures(1)based on a high-throughput theoretical database,machine learning methods are employed to select potential candidates and(2)computational verification is applied to these candidates about their transport *** verification data will be added into the database to improve the extrapolation abilities of the machine learning *** strategies of selecting candidates have been tested,finally the Gradient Boosting Regression model of Query by Committee strategy has the highest extrapolation accuracy(the Pearson R=0.95 on untrained systems).Based on the prediction from the machine learning models,binary pnictides,vacancy,and small atom-containing chalcogenides are predicted to have large *** bonding analysis reveals that the alterations of anionic bonding networks due to small atoms are beneficial to the PFs in these compounds.

关键词： learning bonding prediction

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Study on irradiation hardening of RPV steel based on modified crystal plastic finite element model

Study on irradiation hardening of RPV steel based on modifie...

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2019 Transactions of the American Nuclear Society, ANS 2019

作者： Wang, Xiaotong Yao, Weihua Luo, Ying Dong, Yuanyuan Science and Technology on Reactor System Design Technology Lab Nuclear Power Institute of China Chengdu China Nuclear Power Institute of China Chengdu China

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