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

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Prospects for Next Generation Nuclear Power system 23rd

Prospects for Next Generation Nuclear Power System

引用

23rd Pacific Basin Nuclear Conference, PBNC 2022

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

ISBN: (纸本)9789811988981

systematically research has been carried out on the demand of economic development for low-carbon clean energy and the opportunities faced by the development of nuclear power. A new generation of nuclear power technical index of "Inherent safety, Economics competitiveness, Environment-friendly, and Sustainability" has been formed, which provides guidance for the innovation and development of nuclear power technology. Considering the technical potential of various advanced reactor including the fourth-generation nuclear power system, inherently safe multi-purposed lead-cooled traveling wave reactor is proposed to solve the long-standing problems that have confronted nuclear power. The advanced reactor has attracted much attention because of their unique advantages in fuel supply issue, inherent safety, and economic efficiency. In this study, the main features of the advanced reactors are reviewed and current challenges in their development are pointed out. The advanced reactor obviates the fuel supply issues, including the nuclide enrichment, while making abundant natural uranium usable nuclear fuel. Each of these novel features is technologically separable, so that these new reactor design concepts can be also applied piecewise to enhance prospects in many different utilization circumstances. © 2023, The Author(s), under exclusive license to Springer Nature Singapore Pte Ltd.

关键词： Nuclear energy

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RESEARCH OF MODEL FOR LOCA NONLINEAR DYNAMIC ANALYSIS OF reactor COOLANT system 30

RESEARCH OF MODEL FOR LOCA NONLINEAR DYNAMIC ANALYSIS OF REA...

引用

30th International Congress on Sound and Vibration, ICSV 2024

作者： Yanli, Yuan Xianhui, Ye Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Sichuan Chengdu China

ISBN: (纸本)9789090390581

The Loss of Coolant Accident (LOCA) is one of the most serious accidents in nuclear power plants. According to the requirements of the United States Nuclear Regulatory Commission (USNRC) SPR, the dynamic response analysis of LOCA transients occurring in the reactor coolant system (RCS) must be performed to verify the design of equipment, supports, anti-whip devices and civil structures[1]. At present, the LOCA dynamic response of reactor coolant systems is mainly analysed using a single loop model, and there are few relevant literatures on LOCA analysis with the complete loop model. The influence on the destructive loop due to the hydraulic load of the unbroken loop cannot be considered when using a single loop model for LOCA dynamic response analysis, which may lead to unreasonable analysis results. In this paper, the two models for nonlinear LOCA dynamic analysis of reactor coolant system are established: one is single loop, and the other is three-loop. The nonlinear time-history dynamic analysis of LOCA is analysed by ANSYS software, and the analysis results from the two models are compared. The results show that the loads from the three-loop LOCA dynamic analysis model are consistent with those of the single loop LOCA analysis. By comparing the results calculated by single loop LOCA analysis model with different boundary conditions at inlet nozzles, it is found that the expansion displacements due to the internal pressure at the inlet nozzles of the reactor pressure vessel cannot be ignored during the LOCA dynamic response analysis. © 2024 Proceedings of the International Congress on Sound and Vibration. All rights reserved.

关键词： Nuclear power plants

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Elastic and plastic analysis of O-ring seal for reactor pressure vessel 13th

Elastic and plastic analysis of O-ring seal for reactor pres...

引用

13th Asia-Pacific Symposium on Engineering Plasticity and its Applications, AEPA 2016

作者： Shao, Xuejiao Zheng, Liangang Zhang, Liping Fu, Xiaolong Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610041 China

ISBN: (纸本)9783035710243

The elasto-plastic analysis for reactor Pressure Vessel seal is necessary because of nuclear safety is actively demanded. The existed work based on simplified way to simulate the seal ring, such as uniform stress, spring elements, is too conservative. Therefore, it is necessary to simulate seal ring using 3D solid model considering the elasto-plastc deformation. In this paper, the two-dimensional model is adopted to simulate the compressing and springback character of inconel 718 O-ring using nonlinear isotropic hardening constitutive model and contact algorithm. The simulation approach and constitutive model are validated according to the deformation of seal ring using ANSYS. The seal analysis for reactor pressure vessel is achieved by considering complex loads, such as bolt pre-tightening force and temperature/pressure transient with the usage of an elasto-plastic constitutive model. It is found that the deformation of seal ring obtained by finite element is close to true value. The application of 3D solid model can reduce excessive conservatism effectively and improve the precision of seal analysis compare to the simplified method. © 2017 Trans Tech Publications, Switzerland.

关键词： O rings

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STUDY ON AUTOMATIC DEPRESSURIZATON system OF SMR

STUDY ON AUTOMATIC DEPRESSURIZATON SYSTEM OF SMR

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2020 International Topical Meeting on Advances in Thermal Hydraulics, ATH 2020

作者： Yin, Shasha Sun, Yan Li, Yi Yang, Hongfa Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Sichuan Chengdu China

ISBN: (纸本)9780894487774

Small modular reactor (SMR) is highly cared and researched by people all over the world. There are many differences during reactor operating between the small modular reactor and the traditional large-scale pressurized water reactor, because of the differences in structure and design issues for both reactors. This means that more and better research especially on SMR will be needed. The continued improvements in designing conception of SMR, the research of SMR get to be more in depth and detail. Passive safety system is commonly accepted in the present study, and developed into an indispensable system for the advance nuclear reactor. However, automatic depressurization system (ADS) as an important part of a passive safety system also is a unique system compared with conventional PWR. So ADS system is worthy of comprehensive and in-depth study. In this paper, the severe accident analysis model was established based on the design characteristics of the small modular nuclear. Then, the severe accident process, the thermalhydraulic phenomena and the system safety were researched. Then, based on the results of the severe accident process for SMR, analyze the depressurization capability of ADS system during a station blackout initiated severe accident in SMR, and analysis the effect on accident process. Finally, this paper proposed the optimization and improvement of the automatic depressurization system (ADS). The results of this study are the necessary foundations for the development of small modular reactor severe accident prevention and mitigation measures. © ATH 2020 - International Topical Meeting on Advances in Thermal *** rights reserved.

关键词： Pressurized water reactors

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SMR FEEDWATER system ANALYSIS 30

SMR FEEDWATER SYSTEM ANALYSIS

引用

30th International Conference on Nuclear Engineering, ICONE 2023

作者： Zhu, Ye Liang, Qiao Xianwei, Liao Zhiyun, Cai Minghao, Liu Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Sichuan Chengdu China

ISBN: (纸本)9784888982566

The once through steam generator has great heat transfer capacity and compact structure. Some small module reactors (SMR) applied once through steam generatorfor its high thermal efficiency. The thermal hydraulic characteristic of once through steam generator is different from traditional natural cycle steam generator as it doesn't contain water level. Thus the related feed water system is important for the steady operation of the power plant. Base on the thermal hydraulic characteristic of once through steam generator, SMR feed water system that contains process and control model is developed using APROS. The operating parameters are simulated under both different steady states and transient condition. The results show that both the thermal hydraulic and automatic control parameters of SMR feed water system can be simulated accurately in APROS. Reference and guidance can be provided for system optimization and safety analysis. © 2023 by JSME.

关键词： Water levels

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Response Time Analysis Method of Safety-Level DCS in Nuclear Power Plant Based on Probability Theory 1

引用

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

作者： Wen, Jing He, Xian-Jian Chen, Zhao Wan, Zi-Yuan Wu, Zhi-Qiang Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610000 China

ISBN: (数字)9789819934553

ISBN: (纸本)9789819934546

Aiming at the incompleteness and inadequacy of the response time analysis method of the current nuclear power plant safety-level digital-control system (DCS), this paper proposes an analysis method based on probability theory for the response time of nuclear power plant safety-level DCS. Firstly, according to the structure and characteristics of the safety-level DCS, the probability model of the response time of the safety-level DCS is established, and the calculation method of the probability distribution of the response time is obtained. Then, according to the obtained calculation method and the parameters of the engineering example, the rang of the response time and the probability density function are calculated, and the response time of the engineering example is simulated, and the simulated response time rang, frequency and probability density function fitting curve are obtained. Finally, the response time of the engineering example is tested by the response time automatic test device, a large amount of data with statistical significance is obtained. After the data is processed, the range, frequency and probability density fitting curve of the response time are obtained. By comparing the theoretical calculation results, simulation results and test results of engineering examples. The correctness and feasibility of the analysis method are verified. © 2023, The Author(s), under exclusive license to Springer Nature Singapore Pte Ltd.

关键词： Nuclear power plants

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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 ...

引用

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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Comparisons of Subgrid-Scale Models for OpenFoam Large- Eddy Simulation 7

Comparisons of Subgrid-Scale Models for OpenFoam Large- Eddy...

引用

2020 7th International Conference on Computer-Aided design, Manufacturing, Modeling and Simulation, CDMMS 2020

作者： Feng, Zhipeng Qi, Huanhuan Huang, Xuan Liu, Shuai Liu, Jian Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu Sichuan China

OpenFOAM is a free, open-source software package that can be used for the solutions of computational fluid dynamics and simulation of various fluid flow processes. Nevertheless, OpenFOAM still lacks default settings and a large number of different numerical schemes and turbulent models should be validated. In this paper, the unsteady flow around a cylinder (Re=3900) is calculated by the large eddy simulation of OpenFOAM. The predictions include the drag and lift coefficient, the pressure distribution around the cylinder, the velocity distribution and Reynolds stress distribution in the wake region, as well as the prediction of the recirculation length and separation angle. Thanks to several simulations, these five subgrid-scale (SGS) models are compared and studied: The Smagorinsky SGS model, wall adaptive local eddy visibility SGS model, dynamic SGS model with Lagrangian averaging, dynamic one equation eddy visibility model, one equation eddy visibility model. The numerical results are verified with the published experimental data. © Published under licence by IOP Publishing Ltd.

关键词： Large eddy simulation

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Structural design and Dynamic Analysis of Nuclear Safety DCS Cabinet with Light-Weight and Impact Resistance 5th

Structural Design and Dynamic Analysis of Nuclear Safety DCS...

引用

5th International Symposium on Software Reliability, Industrial Safety, Cyber Security and Physical Protection of Nuclear Power Plant, ISNPP 2020

作者： Dai, Jun-An Liu, Ming-Xing Wu, Xiao Chen, Zhi Wang, Dong-Wei Li, Fa-Qiang Yao, Chang-Wen Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610213 China

ISBN: (纸本)9789811634550

Safety digital control system (DCS) cabinet, as a carrier for the electronic devices, plays a significant role in ensuring the normal operation of the nuclear power plant. In this work, a design idea of impact resistant cabinet frame with high strength and low quality has been proposed. It adopts the design ideas such as flat frame section design, distributed multi-column layout, integrated structural and functional design and preservation of a certain safety margin, which can be used to guide the design of cabinet with high stiffness and strength, small column size and low weight. In addition, finite element analysis and impact response analysis of the designed cabinet are carried out by using ABAQUS-DDAM method. Results show that the designed cabinet is characterized by high stiffness, high strength and low weight, and the cabinet possesses good ability in resisting high level of impact. The design ideas proposed in this research can provide reference for the design of structural parts of other anti-shock electronic equipment. © 2021, The Author(s), under exclusive license to Springer Nature Singapore Pte Ltd.

关键词： Finite element method

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Openfoam Simulation of Damping Controlled Fluidelastic Instability 23rd

Openfoam Simulation of Damping Controlled Fluidelastic Insta...

引用

23rd Pacific Basin Nuclear Conference, PBNC 2022

作者： Feng, Zhipeng Liu, Shuai Qi, Huanhuan Huang, Xuan Lv, Xi Science and Technology On Reactor System Design Technology Laboratory Nuclear Power Institute of China Sichuan Chengdu China

ISBN: (纸本)9789811987793

A parallel triangular tube bundle with a single flexible tube is taken as the object, and the most dangerous mechanism of flow induced vibration in the tube bundle structure, fluidelastic instability (FEI) is studied based on the open-source CFD tool OpenFOAM. Firstly, a numerical model describing the fluid-structure interaction of the tube bundle is established by combining computational fluid dynamics with the vibration equation of the tube;Then, the variation of lift coefficient, drag coefficient, lateral displacement and lift spectrum with velocity is discussed. The results show that the open-source tool OpenFOAM can be easily used to predict the fluidelastic instability of tube bundle. © 2023, The Author(s), under exclusive license to Springer Nature Singapore Pte Ltd.

关键词： Computational fluid dynamics

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