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

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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 ...

引用

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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Fatigue crack growth model for edge sliding mode 14

Fatigue crack growth model for edge sliding mode

引用

14th International Conference on Fracture, ICF 2017

作者： Shi, K.K. Zheng, B. Bai, X.M. Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610213 China

The inherent law between fatigue behaviours of shear-type representative volume element and mode-II fatigue crack growth is found in the range of cycle plastic zone near the crack tip. Prediction models for mode-II fatigue crack growth rate are then proposed by utilizing shear-type low cycle fatigue properties, fatigue failure criterions (plastic strain energy and linear damage accumulation), and effective cycle stress-strain field ahead of the crack tip. Experimental data of two aluminium alloys, 2024-T351 and 7075-T6, are used for models verification. Good agreement between experimental and theoretical results is obtained. © 2017 ICF 2017 - 14th International Conference on Fracture. All rights reserved.

关键词： Fatigue crack propagation

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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...

引用

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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Stress intensity factors for penetrate radial cracks in reactor coolant pump flywheel by weight function method 14

Stress intensity factors for penetrate radial cracks in reac...

引用

14th International Conference on Fracture, ICF 2017

作者： Shi, K.K. Zheng, B. Gao, S.Q. Chen, J.G. Lv, X. Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610213 China

Stress intensity factors for penetrate radial crack at center of keyway and at keyway corner under the plane stress, the plane strain and the three dimensional, based on weight function method, is determined by finite element analysis. Two load cases of pin loading and uniform loading are considered in the present study. About stress intensity factors, finally, an approximate solution and results from weight function method are compared. Several loading are considered that lead to exact solutions which agree with the literature. The accuracy of the analysis has been examined using the finite element method results and were compared to existing solutions for uniform loading in the literature for special geometries, indicating an excellent agreement. © 2017 ICF 2017 - 14th International Conference on Fracture. All rights reserved.

关键词： Stress intensity factors

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Application of optimus in sensitivity analysis of dynamic response of reactor coolant system 27

Application of optimus in sensitivity analysis of dynamic re...

引用

27th International Congress on Sound and Vibration, ICSV 2021

作者： Yanli, Yuan Xianhui, Ye WenZheng, Zhang Lijuan, Li Zhenyu, Liu Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Sichuan Chengdu China

ISBN: (纸本)9788378807995

In this paper, the sensitivity of the dynamic response of reactor coolant system to the input parameters is investigated through OPTIMUS integrated platform, taking the stiffness of the dampers which support the steam generator and the main pump in the reactor coolant system as the input variables. OPTIMUS is a kind of software developed by Noesis Solutions Company, by which the multidisciplinary analysis process can be integrated for collaborative optimization design. OPTIMUS allows users to graphically establish workflows in which they define input files, output files, input parameters, output parameters and analysis tools that are needed to be called and the logical relationships between controls in the process. The dynamic response analysis of reactor system is the key process in the mechanical evaluation. With the continuous improvement of seismic design benchmark and reactor design requirements, the problem of structural stress over-limit is becoming more and more prominent, and the repeated iteration of design to evaluation becomes an inevitable trend. It is necessary to examine the sensitivity of the dynamic response to the key design parameters in the reactor system and find out the design parameters which have a great impact on the reactor dynamic response to shorten the design cycle and improve the design efficiency. In this paper, the sensitivity of the output parameters to the input parameters are investigated by OPTIMUS, considering the stiffness of dampers supporting the steam generator and the main pump as the input parameters, taking the supporting loads of the main equipments as the output parameters. Not only can ANSYS simulation calculations be automatically carried out on the OPTIMUS, but also can the output data be disposed automatically and rapidly. The sensitivity analysis results can be used for the determination of design variables in the subsequent optimization analysis. © "Advances in Acoustics, Noise and Vibration - 2021" Proceedings of

关键词： Sensitivity analysis

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Development of three-dimensional neutron kinetics code based on high order nodal expansion method in hexagonal-Z geometry 26

引用

2018 26th International Conference on Nuclear Engineering, ICONE 2018

作者： Chao, Guo Yu, Liu Hangxing, He Luguo, Liu Xiaoyu, Wang Sufang, Xin Peiying, Li Xiaoli, Wu Hongsheng, Yuan Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610213 China

ISBN: (纸本)9780791851456

To solve three-dimensional kinetics problems, a high order nodal expansion method for hexagonal-z geometry(HONEM) and a Runge-Kutta (RK) method are respectively adopted to deal with the spatial and temporal problem. In the HONEM, 1D partially-integrated flux are approximated by using four order polynomial. The two order polynomial is adopted to the approximation of partially-integrated leakages. The Runge-Kutta method is adopted as a tool for dispersing the time term of 3D kinetics equation. A flux weighting method (FWM) is used for obtaining homogenized cross sections of mix node. The three-dimensional hexagonal kinetics code has been developed based on this method and tested with two benchmark problems of VVER which are the control rod ejection without any feedback and with simple adiabatic Doppler feedback. The results calculated by this code agree well with the reference results and the code is validated. Copyright © 2018 ASME.

关键词： Runge Kutta methods

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Validation of sop procedure for dealing with FWLB 22

引用

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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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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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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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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