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检索条件"机构=Science Technology on Reactor System Design Technology Laboratory"

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Anti-interference Study of Wireless Sensor Networks in Nuclear Power Plants 1

引用

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

作者： Deng, Zhi-guang Wu, Qian Lv, Xin Xiang, Mei-qiong Xu, Tao Qin, Yue Sun, Yong-sheng Peng, Ye-shun Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610213 China

ISBN: (数字)9789819934553

ISBN: (纸本)9789819934546

The compatibility and anti-interference capability of wireless sensor networks within nuclear power plants in complex electromagnetic environments are the most essential elements to ensure reliable operation and effective combat of equipment within nuclear power plants in harsh environments. Interference signals generated by electrical, rotating and communication equipment and sensors attached to nuclear power plants, as well as malicious enemy interference sources, cause their electromagnetic environment to become increasingly complex. For the complex interference environment of nuclear power plant compartment, this paper firstly adopts supervised and unsupervised learning for interference cognition, and after obtaining the features of wireless interference sequences, anti-interference decision based on deep reinforcement learning and array antenna multi-domain anti-interference based on intelligent decision are implemented. From the intelligent interference simulation based on the physical transmission system, it can be seen that the proposed method can effectively identify and resist typical interference signals, which can bring reference to the anti-interference of nuclear power plants. © 2023, The Author(s), under exclusive license to Springer Nature Singapore Pte Ltd.

关键词： Nuclear power plants

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RESEARCH ON LOCA DYNAMIC ANALYSIS MODEL OF reactor system 29

RESEARCH ON LOCA DYNAMIC ANALYSIS MODEL OF REACTOR SYSTEM

引用

2022 29th International Conference on Nuclear Engineering, ICONE 2022

作者： Liu, Shuai Harbin Engineering University Harbin China Science and Technology on Reactor System Design Technology Laboratory Chengdu China

ISBN: (纸本)9780791886472

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 reactor loop system 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 surge 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. Copyright © 2022 by ASME.

关键词： Loss of coolant accidents

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The instrumentation and control system for small modular reactor ACP100 9

The instrumentation and control system for small modular rea...

引用

9th International Topical Meeting on Nuclear Plant Instrumentation, Control, and Human-Machine Interface Technologies, NPIC and HMIT 2015

作者： Chen, Zhi Song, Dan-Rong Liao, Long-Tao You, Kai Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610041 China

ISBN: (纸本)9781510808096

The small modular reactor (SMR), which is coded ACP100 and developed by China National Nuclear Corporation(CNNC), is an innovative PWR based on existing PWR technology, adapting "passive" safety system and "integrated" reactor design technology. It can reach large power by several modular combinations and can be used in different places and different condition. This paper introduces the main technical characteristic of ACPI00 firstly, for example, integrated layout of primary system and equipment, larger primary coolant inventory, once-through steam generator (OTSG), complete passive safety systems, multi-unit layout, and so on. Based on the description of the main technical characteristic of the ACPI00, the basic design rules and requirements for the ACPI00 instrumentation and control (I&C) system are introduced in this paper. The overall I&C system architecture is also introduced. Especially, the design scheme and features of the NSSS I&C systems, including reactor power control system, OTSG feedwater control system, reactor protection system, in-core instrumentation system, ex-core neutron flux monitoring system, are discussed in this paper. Furthermore, the simulation studies for the main control systems are introduced in the paper, including simulation method, simulation models and simulation results. The control scheme of these control systems are validated by the simulation study.

关键词： Pressurized water reactors

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Large eddy simulation of turbulent flow in rod bundle with spacer grids

Large eddy simulation of turbulent flow in rod bundle with s...

引用

2017 Transactions of the American Nuclear Society, ANS 2017

作者： Zonglan, Wei Xiao, Chu Yu, Zhang Songtao, Liu Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610213 China

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design of Code Generator for Safety Level I&C Software of NPPs 6th

Design of Code Generator for Safety Level I&C Software of NP...

引用

6th International Symposium on Software Reliability, Industrial Safety, Cyber Security and Physical Protection of Nuclear Power Plant, ISNPP 2021

作者： Wei, Jiang Lin, Lan Bin, Yang Hou, Rong-Bin Science and Technology On Reactor System Design Technology Laboratory Nuclear Power Institute of the China Sichuan Chengdu China

ISBN: (纸本)9789811911804

The function block diagram is used to design control algorithm of nuclear safety-level I&C (Instrumentation & Control) system. Then it is converted into C code by code generator. And the C code is downloaded into embedded software platform of NASPIC after compiled by a common C compiler. Therefore, the credibility of the code generator is directly related to the safety of nuclear power plant operation. At present, the methods which are used to guarantee the credibility of code generators mainly include test and V&V. However, the formal method describes the software system from a mathematical point of view and verifies its correctness of translation processes, which can improve the credibility of the code generator to the greatest extent. Based on the formal method and compilation technology, this paper introduces the design and implement of code generator for safety I&C software. A initial test has been performed to check the requirement of nuclear I&C software development. The test results show that the code generator can translate Lustre program into C program. © 2022, The Author(s), under exclusive license to Springer Nature Singapore Pte Ltd.

关键词： Nuclear power plants

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Calculation of higher eigenpairs of the transport equation Using IRAM based on domain decomposition

Calculation of higher eigenpairs of the transport equation U...

引用

2017 Transactions of the American Nuclear Society, ANS 2017

作者： Wu, Wenbin Yu, Yingrui Li, Qing Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610041 China

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Research on periodic test scheme of safety digital control system for nuclear power plant 27

Research on periodic test scheme of safety digital control s...

引用

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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Constructing Basic Probability Assignment Based on the Slope Correlation Degree 5

Constructing Basic Probability Assignment Based on the Slope...

引用

5th IEEE International Conference on Electronics technology, ICET 2022

作者： Chen, Jie Yang, Pengcheng Pu, Xiaofei Xie, Hao Wang, Jinxin Wu, Kunren Nuclear Power Institute of China Science and Technology on Reactor System Design Technology Laboratory Chengdu610213 China

ISBN: (纸本)9781665485081

DS evidence theory can be applied to deal with multi-sensor data and get reliable and accurate decision result of system's target or object due to its good measurement and reasonable description of system's uncertainty. In practical application field, DS evidence theory is achieved by two steps, one is the acquisition of the basic probability assignment function (BPA), and the other is the synthesis of multiple evidences that is obtained from multi-sensor data. The existing relevant literatures mainly solve the combination problem of conflict evidences, while lack of systematic research on the acquisition of BPA. In this paper, we focus on the discussion and innovation of the acquisition of BPA. BPA in DS evidence theory is the basic unit to describe the uncertainty of the system and evidences, which fundamentally determines its superiority with respect to probability theory and Bayesian reasoning. Getting a reasonable and effective BPA is the premise of the application of DS evidence theory. Traditional BPA acquisition methods have the limitations like absent priori knowledge and low acquisition accuracy. Aiming to solve the technical difficulties of traditional methods, we put forward a new BPA acquisition algorithm based on the slope correlative degree. The novel algorithm proposed in this paper successfully realizes the reliable descriptions of system's uncertainty, which is conducive to product the accurate BPAs and the precise fusion results. Experimental results and analyses demonstrate that the innovative algorithm can not only produce reasonable BPAs, which can better fit the sensors' monitoring data, but also expand the gap between the support degree of different propositions in BPA, which is more beneficial to the synthesis of evidences and the decision fusion of the system. Thus, the novel BPA acquisition algorithm has great application value and research significance. © 2022 IEEE.

关键词： Uncertainty analysis

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Limit load estimation for finite plates with semi-elliptical defects under tensile loading 9

Limit load estimation for finite plates with semi-elliptical...

引用

9th Global Conference on Materials science and Engineering, CMSE 2020

作者： Shi, K.K. Fu, X.L. Xie, H. Shao, X.J. Tian, J. Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu Sichuan China

The ASME code allows the designer to circumvent many of the problems arising in design by elastic analysis by performing plastic or limit analysis of the component. Unlike elastic analysis, these analyses take account of the redistribution of stress upon yield. In limit analysis, the calculation of the limit load for finite plate with different semi-elliptical defects is an important stage for mechanical safety evaluation. The elastic compensation technique proposed by Mackenzie can be used to provide the limit load of a structure with a finite element method and aims at producing a suitable stress filed for the lower bound theory. In the present study, it is assumed that the finite plate exists three kinds of semi-elliptical cracks with low aspect ratio a/c (a/c=0.1, 0.5 and 1.0), respectively. The improved elastic compensation technique realized by finite element method is applied in the limit load estimation. At last, the results show that the improved elastic compensation technique can be used to obtain the limit value for the finite plate with different semi-elliptical defects under tensile loading. Moreover, the improved elastic compensation technique can be easily implemented in any finite element method and enables the limit load for any complex structure to be obtained swiftly. © Published under licence by IOP Publishing Ltd.

关键词： Finite element method

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Research on Ultra High Flux Research reactor 23rd

Research on Ultra High Flux Research Reactor

引用

23rd Pacific Basin Nuclear Conference, PBNC 2022

作者： Zhang, Xue Yu, Hongxing Xia, Bangyang Li, Wenjie Zhang, Xilin Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610213 China

ISBN: (纸本)9789819910229

The lack of fast neutron irradiation test equipment is a key practical factor that restricts the development of the new generation nuclear energy technology with fast reactor as the core in China. Scarce isotopes such as californium 242, which are urgently needed in industry and medicine, are also highly dependent on imports. In order to provide a fast reactor material irradiation environment and isotope production environment, significantly accelerate the development of fast reactor fuel and structural material technology, and improve the efficiency of rare isotope production, this paper proposes a multifunctional inherently safe ultra-high flux reactor design with fast/thermal neutron flux levels up to 1016 n/(cm2.s), namely UFR-1016. The neutron energy spectrum covers fast spectrum, superheat spectrum and thermal spectrum;several material irradiation test orifices, isotope production orifices and test loop orifices are arranged in the core, and the test loop supports many new reactor coolants such as H2O, Na, He, He-Xr, molten salt, etc., and supports the simulation of typical transient processes and accident conditions. This paper discusses the overall construction scheme of the reactor, and the current construction scheme shows that each performance index can meet the target requirements and achieve the expected functions while ensuring the safety performance of the reactor. This study can provide a basis for the development of ultra-high flux reactors and provide strong support for the cause of nuclear energy, nuclear technology, and nuclear medicine in China. © 2023, The Author(s).

关键词： Nuclear energy

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