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

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Study on control strategy of hydrogen risk in small containment under severe accident 26

Study on control strategy of hydrogen risk in small containm...

引用

2018 26th International Conference on Nuclear Engineering, ICONE 2018

作者： Zou, Zhiqiang Hou, Liqiang Zhang, Ming Deng, Chunrui Peng, Huanhuan Xuedong, Qiao Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China Nuclear and Radiation Safety Center Ministry of Environment Protection of P.R.China Beijing China

ISBN: (纸本)9780791851517

Hydrogen combustion or detonation happened in the containment within the process of the small reactor severe accident may threaten the integrity of the containment. In this paper, based on systemic design of the Small Modular reactor (SMR) surrounded by the steel containment, an innovatory combustible gas control strategy which using the passive containment cooling system (PCCS) and passive autocatalytic recombiners (PARs) is made to control the hydrogen risk in the small steel containment. A severe accident hydrogen risk analysis model is built by the integrative severe accident analysis program MELCOR, the validity of the strategy is analyzed at a typical severe accident. With this understanding, a three-dimensional computed fluid dynamics hydrogen behavior analysis model of the small steel containment is established by GASFLOW code, and the gas distribution non-uniformity in the containment is analyzed. The result shows that the steam condensation process in the containment could be slowed down by controlling the action of PCCS, and the steam concentration in the containment could be in the range of high level, while the oxygen concentration could be in the range of low level. If the PARs were added, the PARs could consume the hydrogen and oxygen in the containment sustainedly. The containment atmosphere could be in an inerted condition during the accident process, even though the hydrogen concentration in the containment is high. The gas distribution non-uniformity analysis result shows that oxygen concentration was low in the extent of high hydrogen concentration and high steam concentration, the steam, oxygen and hydrogen distribution non-uniformity would not affect the inerted atmosphere of containment. Copyright © 2018 ASME.

关键词： Risk analysis

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Semantic Segmentation of Substation Scenes Using Attention-Based Model 4

Semantic Segmentation of Substation Scenes Using Attention-B...

引用

4th IEEE International Conference on Electronics technology, ICET 2021

作者： Chen, Qian Zhang, Changhua Li, Hao Wan, Naijia Wang, Donghui Xu, Bin University of Electronic Science and Technology of China School of Information and Software Engineering Chengdu China University of Electronic Science and Technology of China School of Mechanical and Electrical Engineering Chengdu China Nuclear Power Institute of China Science and Technology on Reactor System Design Technology Laboratory Chengdu China

ISBN: (纸本)9781728176734

Inspection robots are widely used in transformer substation for inspection tasks. In order to complete the task, the robots need to understand and recognize road scenes. However, the transformer substation contains lots of similar and small equipment, which are difficult to recognize. To address this problem, we propose a semantic segmentation model to recognize road scenes. Specifically, our proposed model discriminates the equipment by leveraging multi-view method. We use attention mechanism to strengthen the relationship between pixels that belong to an individual category. Compared with previous methods, our approach achieves a good performance in a new dataset from a transformer substation and shows robustness in recognizing equipment of transformer substation. © 2021 IEEE.

关键词： Roads and streets

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reactor Containment Passive Safety Analysis: Steam Condensation in Presence of Non-condensable Gas Scaled Experiment and Modeling

Reactor Containment Passive Safety Analysis: Steam Condensat...

引用

2024 Pacific Basin Nuclear Conference, PBNC 2024

作者： Bhowmik, Palash K. Schlegel, Joshua P. Sabharwall, Piyush Division of Reactor System Design and Analysis Idaho National Laboratory Idaho FallsID United States Department of Nuclear Engineering and Radiation Science Missouri University of Science and Technology RollaMO United States

ISBN: (纸本)9798331307653

This study presents steam condensation scaled experiments and semi-empirical models in presence of nitrogen (N)—a noncondensable gas (NCG), simulating air in the reactor containment—to support water-cooled small modular reactors (SMRs) passive containment cooling system (PCCS) design and analysis. Previous experimental studies on PCCS are focused on fixed and smaller tube (mostly 2-in.) geometries and specific test condition variations, bringing challenges with geometric scaling and mismatching with SMR prototypic design. To address these challenges, this study presents steam condensation test dataset obtained from three scaled test sections of 1-, 2-, and 4-in.-diameter steam condensers with an annular/ jacket cooling of 2-, 3-, and 6 in.-diameter tubes, respectively. Test data were collected for steam ranges from 58 to 63 kg/hr., and NCG flow of to 13.3 kg/hr. Annular cooling water flow was varied to obtain required testing conditions of saturated steam 4.4 inlet and fully condensed outlet. Axial temperature test data of bulk cooling water, steam and condensate were collected by thermocouples for three test sections and various steam-NCG mixing/testing conditions. A standard data reduction method was adopted—utilizing iterative and nodalized mass and heat transfer calculation—to estimate axial local heat fluxes, heat transfer coefficients (HTCs), condensation rates, film thickness, and Nusselt number. Based on the obtained dataset semi-empirical model results—a ratio of experimental and Nusselt’s theoretical HTC are presented. Such results and findings are supportive of developing scaled-up testing facility, to model validations and accelerate next generation of reactors development and deployment. © 2024 Pacific Basin Nuclear Conference, PBNC 2024. All rights reserved.

关键词： Thermocouples

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Microstructure and mechanical properties of tib2/ al-si composites fabricated by electron beam rapid manufacture 21st

Microstructure and mechanical properties of tib2/ al-si comp...

引用

21st Chinese Materials Conference, CMC 2020

作者： Yang, Qingfeng Xia, Cunjuan Li, Xianfeng Wang, Haowei Liao, Nan Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610213 China The State Key Laboratory of Metal Matrix Composites Shanghai Jiaotong University Shanghai200240 China

ISBN: (纸本)9783035717389

Bulky sample was fabricated by electron beam rapid manufacture (EBRM) technology, in which Ф1.6 mm wire of in-situ TiB2/Al-Si compositeswas selected as deposition metal, following byT6 heat treatment. The microstructure and mechanical properties of the bulky sample before and afterheat treatment were analyzed. Experimental results showed that the microstructure parallel to the weld was similar to that perpendicular to the weld. The microstructure of the as-deposited sample consisted of columnar and equiaxed grains, in which siliconwas distributed along the grain boundary and the grain size was about 30 μm. Besides, some TiB2 particles converged at the grain boundary. After T6 heat treatment, the average grain size of the sample *** average hardness of the sample was increased to 114.65 HV from 46.55 HV, an increase of 146%. The tensile strength of the sample increased to 326.66MPafrom 143.97 MPa, but the elongationdecreased compared with that of the as-deposited state. The tensile test showed that the mechanical properties of TiB2/Al-Si composites formed by electron beam rapid manufacture were isotropic before and after heat treatment. © 2021 Trans Tech Publications Ltd, Switzerland.

关键词： Electron beams

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Multi-physics Coupling Analyses of Nuclear Thermal Propulsion reactor 23rd

Multi-physics Coupling Analyses of Nuclear Thermal Propulsio...

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23rd Pacific Basin Nuclear Conference, PBNC 2022

作者： Han, Wenbin Guan, Zechuan Huang, Shanfang Deng, Jian Department of Engineering Physics Tsinghua University Beijing China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China China Institute of Atomic Energy Beijing China

ISBN: (纸本)9789819910229

Nuclear thermal propulsion (NTP) reactors have high-temperature solid-state characteristics and significant thermal expansion, which therefore require multi-physics coupling analyses. In this paper, the framework of Neutronics, Thermal-Hydraulics and Mechanics coupling (N/T-H/M) of nuclear thermal propulsion reactor is developed, and the typical reactor XE-2 is analyzed with this method. The results show that the N/T-H/M coupling will bring -1049 pcm negative reactivity, of which the thermal expansion effect accounts for 22%, indicating that the nuclear thermal propulsion reactor has a certain capacity for self-regulation. However, thermal expansion will lead to 0.88 mm peak deformation and 233 MPa peak stress, which will severely threaten the mechanical tolerance of the materials. Therefore, there is a trade-off between the advantages and disadvantages of the high-temperature solid-state core while designing NTP reactors. © 2023, The Author(s).

关键词： Thermal expansion

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DEVELOPMENT OF QUENCH ANALYSIS PROGRAM BASED ON ARSAC 29

DEVELOPMENT OF QUENCH ANALYSIS PROGRAM BASED ON ARSAC

引用

2022 29th International Conference on Nuclear Engineering, ICONE 2022

作者： Zhong, Lei Yu, Hongxing Chen, Deqi Liu, Hanzhou Deng, Jian Wu, Dan Science and Technology on Reactor System Design Technology Laboratory Sichuan Chengdu610041 China Key Laboratory of Low-grade Energy Utilization Technologies and Systems Chongqing University Ministry of Education Chongqing400044 China

ISBN: (纸本)9780791886410

ARSAC is a reactor system analysis program independently developed in China. In this paper, a quench analysis module is developed based on ARSAC program. The boiling curve suitable for quench process is established and several classical minimum film boiling temperature and models are implanted. The new program can reflect the influence of minimum film boiling temperature and heat transfer models on the quench process, thus it can be used to estimate the effect of surface oxidation and different cladding materials on the quenching process. The calculation results of the new program are in good agreement with the experimental data, and can better simulate the quench process of fuel element under LOCA. Copyright © 2022 by ASME.

关键词： Heat transfer

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Development of self-healing coating on fuel cladding for water-cooling nuclear reactors 14

Development of self-healing coating on fuel cladding for wat...

引用

14th International Nuclear Fuel Cycle Conference, GLOBAL 2019 and Light Water reactor Fuel Performance Conference, TOP FUEL 2019

作者： Duan, Z.G. Xin, G.S. Zhou, Y. Li, Y.M. Abe, H. Yang, H.L. Kano, S. McGrady, J. Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610041 China Nuclear Professional School School of Engineering University of Tokyo Tokai Ibaraki319-1188 Japan

A novel scheme for a bilayer coating with self-healing ability is proposed in this study. The candidate materials for the coatings and the potential self-healing reaction are assessed in high-temperature aqueous environments and high-temperature air. The pure Cr2O3 layer and the composite of Cr2O3 and MoO3 are the candidate materials for the outer layer and inner layer, respectively, due to their compatibility under normal condition. Fe2O3-MoO3 reactions exhibit a potential ability to heal the cracks because of a high reaction rate under normal condition. Copyright © GLOBAL 2019 - International Nuclear Fuel Cycle Conference and TOP FUEL 2019 - Light Water reactor Fuel Performance *** rights reserved.

关键词： Light water reactors

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Stratification and heat transfer of molten corium pool for in-vessel retention

Stratification and heat transfer of molten corium pool for i...

引用

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

作者： Xiang, Qingan Deng, Jian Zhu, Dahuan Wu, Xiaoli Bi, Jinsheng Chen, Baowen Cai, Rong Qian, Libo YugaoMa Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China Nuclear and radiation safety center Ministry of Environment Protection of P.R. China Beijing China

ISBN: (纸本)9780791883778

In-vessel retention (IVR) consists in cooling the molten corium contained in the lower head of reactor vessel by natural convection and reactor cavity flooding. The general approach which is used to study IVR problems is a "bounding" approach which consists in assuming a specified corium pool stratification in the vessel and then demonstrating that the vessel can cope with the resulting thermal and mechanical loads. Thermal loading on the vessel is controlled by the convective heat transfer inside the molten corium pool. Traditionally molten corium pool in the lower head was expected to stratify into two-layer with the dense oxide pool at the bottom and the light metal layer on the top. Based on the MASCA experiments, the increased density of the metal layer attributed to a transfer of uranium metal leads to inverse stratification with a heavy metal layer relocating below the oxide pool. This behavior can be explained by physicochemical interaction between the oxidic and metallic phases of the corium pool. Therefore, a methodology which couples physicochemical effects and thermal hydraulics has been developed to address the IVR issue. The main purpose of this paper is to present this methodology and its application for calculate stratification probability of two-layer and three-layer configuration, analyze the safety margin of IVR in two-layer and three-layer configuration, and evaluate the lower head heat thermal failure probability. Copyright © 2020 ASME

关键词： Heat flux

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Disorders in fluid filled pipeline structure with elastic helmholtz resonators 29

Disorders in fluid filled pipeline structure with elastic he...

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2022 29th International Conference on Nuclear Engineering, ICONE 2022

作者： Zeng, Qingna Wang, Donghui Zang, Fenggang Zhang, Yixiong Wang, Bihao Yuan, Zhihao Science and Technology on Reactor System Design Technology Laboratory Chengdu610213 China Nuclear Power Institute of China Chengdu610213 China College of Shipbuilding Engineering Harbin Engineering University Harbin150001 China

ISBN: (纸本)9780791886359

This paper studies the influence of hybrid disorders on acoustic transmission characteristic in pipeline with periodic elastic HRs. Band gap structures (BGs) for infinite period are investigated by transfer matrix method and finite element method respectively, and fit well with the corresponding transmission loss (TL) for finite regular periods. By introducing a certain degree of uncertainty level, Interval analysis method is used to convert uncertain parameter problems into two deterministic models. Frequency response function (FRF) for near-periodic structure with single disorder and BGs for quasiperiodic structure with multiple disorders considering cavity radius and periodic length are investigated. The existence of disorders in periodic structure would always reduce attenuation interval and intensity, some disturbance would even change the original acoustic transmission characteristic and excite completely new attenuation zones. The sensitivity of pipeline acoustic vibration transfer characteristic to disorder parameter depends on attenuation mechanism, as for periodic elastic HRs pipeline, acoustic propagation stability is more sensitive to characteristic of locally resonant unit. The work in this paper enriches the stability analysis of pipeline with elastic HRs, which would provide theoretical reference for noise and vibration reduction design in practical engineering applications. © 2022 American Society of Mechanical Engineers (ASME). All rights reserved.

关键词： Transfer matrix method

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Research and Application of Nuclear reactor Computational Data Framework Based on HDF5 7th

Research and Application of Nuclear Reactor Computational Da...

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7th International Conference on Artificial Intelligence and Security, ICAIS 2021

作者： Lu, Wei Feng, Jintao Yang, Hao Zeng, Hui Pan, Junjie Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610213 China School of Computer Science Chengdu University of Information Technology Chengdu610225 China School of Information and Software Engineering University of Electronic Science and Technology of China Chengdu610054 China

ISBN: (纸本)9783030786175

Nuclear reactor computing software need to process and maintain complex and massive data sets. In order to meet the requirement of mass data storage and processing in software, the data storage model and I/O method and programming interface of HDF5 (Hierarchical Data Format v5) are deeply studied. According to the data storage and processing requirement of typical nuclear reactor computing data such as assembly or core neutronics computing data and core thermal-hydraulic computing data, RCDF-H5 (reactor Compute Data Framework based on HDF5) is proposed. KYMRES (KYlin-2 Main RESults databank) and COMRES (COre Main RESults databank) are designed and implemented based on RCDF-H5. The performance tests show that RCDF-H5 has a higher I/O efficiency than conventional storage solutions. A new data storage and management solution for nuclear reactor computing software is provided. © 2021, Springer Nature Switzerland AG.

关键词： Digital storage

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