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

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Hydrogen risk analysis of a large pressurized water reactor using hydragon code 27

Hydrogen risk analysis of a large pressurized water reactor ...

引用

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

作者： Zhang, Xiaolong Yu, Jiyang Huang, Tao Jiang, Guangming Zou, Zhiqiang Department of Engineering Physics Tsinghua University Beijing100084 China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China

ISBN: (纸本)9784888983051

This paper presents the hydrogen risk analysis of a large Pressurized Water reactor (PWR) undergoing a severe accident using the HYDRAGON code. The code has been developed by the Department of Engineering Physics, Tsinghua University, as a hydrogen risk analysis software specialized for containment building of nuclear reactor power plants. The chemical interaction between zirconium cladding of the fuel rods and overheated coolant may release hydrogen when severe accidents occur in PWRs. The hydrogen mixed with air existing in containment may jeopardize the integrity of containment building if the flammable mixture explodes. In order to mitigate the hydrogen risk, many nuclear power plants install the Passive Autocatalytic Recombiners (PARs) that allow hydrogen and oxygen react chemically at low temperature without any sights of flame. Recombiners remove hydrogen and thereby prevent explosions. This work simulated the gas diffusion behaviors during the accident in the containment of a large PWR. The simulations obtained spatial concentration distribution of various gas species including hydrogen, oxygen, steam and nitrogen and several key parameters that are required to complete hydrogen burning and explosion investigation. Sensitivity analyses of recombiners' quantity and positions and hydrogen discharge intensity provides the reference for hydrogen mitigation measure design in the containment. The results demonstrate that the number of recombiners rather than positions has stronger influences on hydrogen mitigation. Since the installation position reflects less sensitivity, the recombiners could be arranged as higher as possible if necessary. Copyright © 2018-2019 by JSME

关键词： Hydrogen

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Study on 131I Source Term in reactor Coolant Under Shutdown Transient Condition 23rd

Study on 131I Source Term in Reactor Coolant Under Shutdown ...

引用

23rd Pacific Basin Nuclear Conference, PBNC 2022

作者： Fu-ting, Jing Huan-wen, L.V. Ming-Ming, X.I.A. Zhang-Han, Y.E. Xi-long, Gao Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China Fuqing Nuclear Power Co. Ltd Fuqing China

ISBN: (纸本)9789811988981

131I activity in coolant under transient conditions is an important input for SGTR accident source term analysis in nuclear power plant, which directly affects whether the radioactive consequences of SGTR can meet the regulatory requirements. Based on the measurement results of 131I source term of several units in nuclear power plant A during shutdown, this paper study the variation law of 131I source term in the transient process of shutdown, analyzes the causes of 131I peak release, calculates the 131I release rate in the transient process, and obtains the following conclusions: 1) the 131I release rate based on the measured value is significantly lower than that used in the current safety analysis;2) The larger the fuel rod break size, the smaller the peak factor of 131I in the primary circuit during transient. © 2023, The Author(s), under exclusive license to Springer Nature Singapore Pte Ltd.

关键词： Nuclear power plants

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CFD Analysis on Hydrogen Distribution in Subcompartment of 600MW NPP under Severe Accident

CFD Analysis on Hydrogen Distribution in Subcompartment of 6...

引用

The Fifth China-Korea Workshop on Nuclear reactor Thermal Hydraulics

作者： ZOU Zhi-qiang ZHANG Yu DENG Jian Science and Technology on Reactor System Design Technology Laboratory

Hydrogen combustion in the containment building may be a threat to containment *** in some subcompartments,hydrogen detonation is more easier to happen,which may destroy the containment structure or the safety equipments in the containment. A study has been carried out using the GASFLOW three-dimensional CFD code to evaluate the hydrogen distribution in the subcompartment during a severe *** Chinese 600MW NPP containment is modeling by GASFLOW. The GASFLOW calculation gave detailed results for the spatial distribution of gas concentrations in the reactor coolant pump（RCP） compartment and the steam generator（SG） compartment within containment.

关键词： Subcompartment Hydrogen LBLOCA+ failure of ECCS CFD

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A Unified Framework of Stabilized Finite Element Method for Solving the Boltzmann Transport Equation

A Unified Framework of Stabilized Finite Element Method for ...

引用

2022 International Conference on Physics of reactors, PHYSOR 2022

作者： Cao, Liangzhi Fang, Chao Li, Yifei He, Qingming School of Nuclear Science and Technology Xi'an Jiaotong University Shaanxi Xi'an710049 China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610041 China

ISBN: (纸本)9780894487873

This paper presents a unified framework of stabilized finite element method for solving the Boltzmann transport equation. Unlike the traditional Petrov-Galerkin finite element method which modify the test function to construct the stabilization term, we derive the stabilization methods from the standard Galerkin weak form with Sub-grid scale model. The basic idea of this method is to decompose the unknowns into its numerical solution and residual, with an approximation for the residual and embed it in the Galerkin weak form yields a stabilized variational formula. Different approximations of the residual lead to different stabilization methods, all the frequently used stabilized methods, including the Streamline Upwinding Petrov-Galerkin (SUPG) method, Galerkin/Least-Square (GLS) method, and Algebraic Sub-Grid Scale (ASGS) method can be obtained from this framework. The similarities and differences of the different approximations are compared in this paper. The numerical results show that the behaviors of the different methods are similar with the same stabilization parameter, and all these stabilized techniques can obtain correct and stable solution. © 2022 Proceedings of the International Conference on Physics of reactors, PHYSOR 2022. All Rights Reserved.

关键词： Approximate residual equation Boltzmann transport equation Stabilization parameter Stabilized finite element methods

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The application of Deep Reinforcement Learning in Coordinated Control of Nuclear reactors 4

The application of Deep Reinforcement Learning in Coordinate...

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2021 4th International Conference on Mechatronics and Computer technology Engineering, MCTE 2021

作者： Li, Jing Liu, Yanyang Qing, Xianguo Xiao, Kai Zhang, Ying Yang, Pengcheng Yang, Yue Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610213 China Department of Engineering The University of Sydney Sydney2006 Australia

The nuclear reactor control system plays a crucial role in the operation of nuclear power plants. The coordinated control of power control and steam generator level control has become one of the most important control problems in these systems. In this paper, we propose a mathematical model of the coordinated control system, and then transform it into a reinforcement learning model and develop a deep reinforcement learning control algorithm so-called DDPG algorithm to solve the problem. Through simulation experiments, our proposed algorithm has shown an extremely remarkable control performance. © 2021 Institute of Physics Publishing. All rights reserved.

关键词： Steam generators

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Single event effect evaluation of different communication ways based on 40nm SRAM-FPGA 3

Single event effect evaluation of different communication wa...

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3rd International Conference on Artificial Intelligence and Computer Engineering, ICAICE 2022

作者： Zhao, Xu Du, Xuecheng Xiong, Xu Chen, Zhi Jiang, Wei School of Nuclear Science and Technology University of South China Hengyang City421001 China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610213 China

ISBN: (纸本)9781510663473

Field-programmable gate array (FPGA) is now considered by an increasing number of designers in various fields of application, such as space and aircraft embedded control systems, image and signal processing. It is a popular way to design different serial communication protocols including serial and network communications implemented by FPGA, which ensure data transfer accuracy and timing. However, single event effects (SEE) can pose a serious threat to the reliability of FPGA in the radiation environments. In order to evaluate the reliability and failure modes of different communication interfaces in radiation environments, three communication protocols were tested by alpha radiation source, including universal asynchronous receiver/transmitter (UART), ethernet, and universal serial bus (USB). The experiments have presented the failure modes and SEE cross sections to different communication ways. © 2023 SPIE.

关键词： Data transfer

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Experimental investigation on temperature characteristics of direct contact condensation in a natural circulation system

Experimental investigation on temperature characteristics of...

引用

2020 International Conference on Nuclear Engineering, ICONE 2020, collocated with the ASME 2020 Power Conference

作者： Sun, Jianchuang Ding, Ming Mi, Zhengpeng Zhang, Zhuohua Fundamental Science on Nuclear Safety and Simulation Technology Laboratory Harbin Engineering University Harbin China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China

ISBN: (纸本)9780791883785

Direct contact condensation (DCC) is a common physical phenomenon appearing in the nuclear power plants and other industrial applications. The current research on DCC focuses on steam-water counterflow or jet flow under forced flow conditions. For some natural-circulation passive safety systems in floating nuclear power plants, the heated section is connected with the heat sink by the horizontal pipes. The heat sink is usually the ocean. In such natural circulation systems (NCSs), the steam produced in the heated section will inevitably contact the subcooled water in the ocean and result in DCC event. In this paper, the fundamental characteristics of two-phase flow were described in detail. In addition, the influences of the subcooled water temperature on the flow rate, outlet temperature, steam bubble behaviors, and pressure surge were emphatically discussed. The experimental results revealed that the subcooled water in the water tank could reversely flow into the pipe, which was able to cause condensation induced water hammer (CIWH) event and flow oscillation. Further research showed that the subcooled water was evidently prevented from reversely flowing into the pipe with the increase in the subcooled water temperature. The position where the DCC event occurs transferred from the pipe to the water tank, and the pressure peak obviously reduced, even disappear when the subcooled water temperature is larger than 61 oC. © 2020 American Society of Mechanical Engineers (ASME). All rights reserved.

关键词： Water temperature

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Research on probabilistic safety assessment (PSA) method of safety class distributed control system (DCS) in nuclear power plant based on fuzzy Markov process

Research on probabilistic safety assessment (PSA) method of ...

引用

2020 International Conference on Nuclear Engineering, ICONE 2020, collocated with the ASME 2020 Power Conference

作者： Tian, Xufeng Hao, Yan Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China HuaFu avenue 299# section 3 Chengdu city610213 China

ISBN: (纸本)9780791883778

Safety class DCS(Distributed Control system) in nuclear power plant is characterized by multi-state interaction, complex function and unknown fault mechanism, which limits the use of tradition PSA(Probabilistic Safety Assessment) method such as event tree and fault tree. An improved dynamic probabilistic security simulation method is proposed based on fuzzy Markov process to quantitative evaluation the risk of Safety class DCS in nuclear power plant, which resolved the failure uncertainty problem without in-depth understanding of failure mechanism. Firstly, the fuzzy fault tree reliability evaluation model is established based on fuzzy theory and FTA(Fault Tree Analysis), which describes the probability and degree of failure with fuzzy number. Then, the fuzzy fault tree is transformed into fuzzy Markov chain, this model is described mathematically, and the reliability indexes are simulated and analyzed with Monte Carlo simulation method, the change of system safety performance in long-term operation is obtained. Finally, the scientificity and effectiveness is verified by the example of redundant analog output module of safety class DCS in nuclear plant. Copyright © 2020 ASME

关键词： Nuclear power plants

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Analysis and verification of burnup algorithm in reactor 27

Analysis and verification of burnup algorithm in reactor

引用

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

作者： Hu, Yuying Luo, Qi Yao, Dong Yu, Yingrui Liao, Hongkuan Zhou, Bingyan Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu Sichuan China Nuclear Power Institute of China Chengdu Sichuan China

ISBN: (纸本)9784888983051

Burnup algorithm analysis is an important part of the physical calculation in reactor. The computing accuracy of burnup algorithms is directly related to the correctness of the whole physical calculation. Due to the difference from the rates and lives of varieties of nuclides, burnup matrices are always sparse rigid exponential ones. At present, methods to solve exponential matrices are divided into two types, numerical methods and analytical methods. Besides, numerical methods conclude series methods, ordinary differential equation methods, polynomial methods, matrix decomposition methods and so on. In this study, some common methods to solve burnup matrices are introduced, such as Chebyshev Rational Approximation Method(CRAM), Krylov Subspace Method, Pade Approximation Method and Transmutation Trajectory Analysis(TTA). Analysis and verification for microscopic point-depletion calculation are carried out with two typical burnup chains. One of the burnup chains for verification is a decay chain of neptunium-237(237Np), which contains 13 nuclides. The other chain is an actinide nuclide reaction chain, including 25 nuclides from decay and neutron capture and 9 nuclides from fission. Because there is no close annual chain in the burnup chains used in verification, the results of using Transmutation Trajectory Analysis were thought to be accurate, and it has been used as the benchmark for comparative analysis of other methods. The final result shows, Chebyshev Rational Approximation Method has advantages in calculation accuracy and simplicity, and it is suitable to be used in large burnup calculations. Copyright © 2018-2019 by JSME

关键词： Isotopes

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Numerical investigations of free surface vortices at hydraulic bottom intakes 17

Numerical investigations of free surface vortices at hydraul...

引用

17th International Topical Meeting on Nuclear reactor Thermal Hydraulics, NURETH 2017

作者： Shi-wen, Liu Jie, Zhou Wei, Huang Xiang, Cheng Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China 328# Section 1 Changshun Road ShuangliuChengdu610041 China

To study the development mechanisms and flow structure characteristics of the free surface vortices and get a better understanding of the vortices phenomena. Two-phase, three dimensions, transient CFD simulations of free surface vortices at hydraulic bottom intakes have been conducted by using a volume-of-fluid (VOF) method and the scale-adaptive simulation turbulence model with curvature correction (SAS-CC). The flow field distribution and structure in different stages of transient free surface vortices were presented and analyzed. It was found that the movement and development of the vortex was determined by the interactions of main flow transportation, the axial stretching and radial dissipation of vortex. In addition, the flow characteristics on different flow conditions also have been studied, the results showed that a relatively large flow rate and small export size lead to stronger vortex, and increasing the fluid viscosity would to some extent reduce the vortex. © 2016 Association for Computing Machinery Inc. All Rights Reserved.

关键词： Vortex flow

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