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Validation of sop procedure for dealing with FWLB 22

2014 22nd International Conference on Nuclear Engineering, ICONE 2014

作者： Yaou, Shen Wei, Chen Daishun, Huang Guijun, Leng Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu Sichuan China

ISBN: (纸本)9780791845936

The State Oriented Procedure (SOP) is developed to improve the existing emergency operation procedures, namely, the Event Oriented Procedure (EOP). It is developed based on the finite Nuclear Steam Supply system (NSSS) states instead of an infinite number of possible events, therefore it can cover all of the events/accidents including the accidents caused by more than one initial event. On the other hand, the operator can take actions without diagnosing the events/accidents according to SOP, and thus the human mistakes which will possible happen during the events/accidents diagnosing can be avoided. However, as a generalized procedure system, the SOP needs a lot of validations based on various events/accidents. In this work, an analysis of a postulated Feed Water Line Break (FWLB) accident for LingAo phase II NPP based on the SOP is given and both the automatic protections and operator manual actions required by the SOP are modeled with an advanced, two-fluid, thermal-hydraulic code called CATHARE. The initial conditions and assumptions are chosen to get a severe state of the NPP. Since the EOP addresses the given accident perfectly, the results are then compared with those based on the EOP for a similar NPP to validate the SOP. The main parameters relevant with FWLB of this NPP are the same with LingAo phase II. Therefore, there is no need to perform analyses based on the EOP for LingAo phase II. The results demonstrate that the main operator manual actions required by the SOP are similar to those required by the EOP, and the operator can take the plant to a controllable safe state according to the SOP procedures. Therefore, the SOP is effective and reasonable for FWLB accident. Copyright © 2014 by ASME.

关键词： Nuclear engineering

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

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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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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...

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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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RESEARCH ON LOCA DYNAMIC ANALYSIS MODEL OF reactor system 29

RESEARCH ON LOCA DYNAMIC ANALYSIS MODEL OF REACTOR SYSTEM

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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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Interfacial ion regulation on 2D layered double hydroxide nanosheets for enhanced thermal insulation

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science China Chemistry 2022年第5期65卷 898-904页

作者： Chun Wang Hengyu Xu Han Cheng Hao Yu Si Liu Wenjie Wang Ruilin Yuan Hongfei Liu Tianpei Zhou Wangsheng Chu HengAn Wu Yi Xie Changzheng Wu Hefei National Laboratory for Physical Sciences at the Microscale University of Science and Technology of ChinaCAS Center for Excellence in Nanoscienceand CAS Key Laboratory of Mechanical Behavior and Design of MaterialsHefei 230026China CAS Key Laboratory of Mechanical Behavior and Design of Materials Department of Modern MechanicsCAS Center for Excellence in Complex System Mechanics University of Science and Technology of ChinaHefei 230027China National Synchrotron Radiation Laboratory University of Science and Technology of ChinaHefei 230029China Institute of Energy Hefei Comprehensive National Science CenterHefei 230031China

Exploring new strategies to broaden the upper/lower limit of thermal conductivity is of great interest to develop thermal management materials that can adapt to extreme *** this work,we employ an interfacial ion regulation to enhance the thermal insulation performance of 2D layered double hydroxide *** introduction of interfacial ion enlarges the interplanar spacing of Co(OH)_(2) nanosheets from 4.64 to 8.05 ?,which reduces phonon scattering length perpendicular to the two-dimensional plane and leads to enhanced interlayer thermal *** interfacial ion-regulated Co(OH)_(2)(named as Co(OH)_(2)-M^(x-)) exhibits 3-fold enhancement of thermal insulation through decreasing the thermal conductivity to as low as 0.15 W m^(-1) K^(-1),which is among the top values in 2D solid *** anticipate that interfacial ion regulation for 2D nanosheets paves a new avenue to break through the thermal insulation limit.

关键词： two dimensional materials interfacial ion regulation thermal conductivity phonon scattering

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The instrumentation and control system for small modular reactor ACP100 9

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

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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...

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

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

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

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

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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

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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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