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

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Research on the influence of ocean condition on the oscillation in parallel double-channel

Research on the influence of ocean condition on the oscillat...

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18th International Conference on Nuclear Engineering, ICONE18

作者： Zhou, Linglan Hong, Zhang Department of Physic Engineering Tsinghua University China National Key Laboratory of Science and Technology on Reactor System Design Technology Nuclear Power Institute Of China China

ISBN: (纸本)9780791849323

In this paper, the flow oscillation in the parallel multichannel system has been studied under ocean condition with RELAP5/MC. A double-channel boiling system model is built using the code RELAP5/MC and it is proved to be capable of simulating the instable phenomena. The influences of rolling, heaving and inclination on the flow oscillation have been analyzed. The result shows that the effect of lengthways rolling on the oscillation in parallel channels can be ignored. The influence of athwartships rolling and heaving reveals resonance effect, that is, when the rolling cycle is close to the oscillation cycle and half cycle respectively, the oscillation would happen earlier, when the difference between the cycles is large, the impact of rolling can also be ignored. The influence of inclination on the oscillation in parallel channels is feeble. © 2010 by ASME.

关键词： Oscillating flow

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Research on trans critical characteristic on sliding pressure booting process of SCWR

Research on trans critical characteristic on sliding pressur...

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2013 21st International Conference on Nuclear Engineering, ICONE 2013

作者： Liu, Liang Zhou, Tao Song, Mingqiang Chen, Juan Luo, Feng Xia, Bangyang North China Electric Power University Beijing China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China

ISBN: (纸本)9780791855836

The homogeneous flow model of trans critical transient calculation is achieved by the trans critical program of SCWR which is made by FORTRAN. The SCWR sliding pressure starting scheme is simulated by this program. So that the pressure raising stage characteristic of its booting process is analyzed. The calculation result shows that, in the pressure raising stage of booting process, the maximum cladding temperature gradually increases. Coolant temperature increases gradually along with the flow direction. The dry degree also increases along with the flow direction. The highest temperature of cladding does not exceed 650 °C limits. There is no major adverse effect on heat transfer of fuel rod in the dramatic changes in the physical properties of water. Copyright © 2013 by ASME.

关键词： Heat transfer

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Preliminary analysis of FCM fuel thermal-mechanical performance under normal and transient conditions

Preliminary analysis of FCM fuel thermal-mechanical performa...

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Top Fuel 2016: LWR Fuels with Enhanced Safety and Performance

作者： Li, Wenjie Tang, Changbing Liu, Zhenhai Li, Wei Chen, Ping Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610213 China Design Sub-institute Nuclear Power Institute of China Chengdu610213 China

ISBN: (纸本)9780894487309

FCM fuel is a candidate fuel form to replace commercial UO2 fuel rod in LWR to enhance fuel reliability under accident conditions. The FCM pellet consists of TRISO particles embedded inside the dense SiC matrix. This fuel design is more complex than conventional LWR fuel, which needs new fuel performance models or analysis tools to assist fuel design and concept evaluation. In this paper an attempt was made to develop a finite element method to analyze FCM fuel pellet thermal-mechanical performance. Three idealized cases of normal operation, power ramp, and LOCA were analyzed. The results showed that the temperature distribution and stress distribution of FCM pellet exhibited a strong non-uniformity. The irradiation strain and creep of buffer layer and PyC layers had a strong effect on the mechanical performance of FCM pellet. The hoop stresses of SiC layers and SiC matrix increased significantly during the LOCA and varied with radial positions.

关键词： Finite element method

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Intra Fast Mode Decision Algorithm and Hardware design for AVS2

Intra Fast Mode Decision Algorithm and Hardware Design for A...

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2021 International Conference on Signal Processing and Machine Learning, CONF-SPML 2021

作者： Xie, Hao Liu, Kaiyang Zhao, Yang Zheng, Xiao Qing, Xianguo Nuclear Power Institute of China Science and Technology on Reactor System Design Technology Laboratory Chengdu China Peking University School of Software and Microelectronics Beijing China

ISBN: (纸本)9781665417341

AVS2 is a kind of video coding standard proposed by China, it adopts 33 intra prediction modes to improve coding performance, while the computational complexity has increased dramatically. In order to reduce the complexity of AVS2 intra mode decision and make the AVS2 hardware encoder meet real-time requirements, this paper proposes the AVS2 intra fast mode decision algorithm and hardware design. The experimental results show that the intra mode decision algorithm and hardware design proposed in this paper can meet the throughput requirement of 1920x1080@60fps at a clock frequency of 300MHz. By using Xilinx FPGA XC7K325T 900 on the Vivado HLS platform for synthesis, only 10% of the LUT, 5% of FF, 5% of BRAM, and 6% of DSP in FPGA resources are consumed to meet the design requirements. © 2021 IEEE.

关键词： Field programmable gate arrays (FPGA)

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Calculation strategy for the determination of plasticity correction factors 13th

Calculation strategy for the determination of plasticity cor...

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

作者： Du, Juan Shao, Xue Jiao Kan, Qian Hua Zhang, Ying Fu, Xiao Long Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610041 China Southwest Jiaotong University Chengdu610031 China

ISBN: (纸本)9783035710243

This paper presents an investigation about the plasticity correction factor, Ke, proposed in the RCC-M code, for the correction of elastic stress range exceeds twice the yield stress which results from both mechanical and thermal loading. The plasticity correction factors are analyzed using a calculation strategy based on the comparison between linear elastic and elastoplastic analyses. The RCC-M code provides the codified expressions of Ke and material parameters for general materials of components in nuclear power plant, but there are no expressions and material parameters for new materials. Several elastoplastic models, geometries and loading types are considered in this study based on the general materials listed in RCC-M code to determine the calculation strategy of Ke for new material. Based on the results, the expressions and parameters of Ke for titanium alloy are discussed. It turns the proposed Ke factor of titanium alloy can yield to sufficiently conservative results. A verification calculation of general materials is performed and the conservative allowance is obtained. © 2017 Trans Tech Publications, Switzerland.

关键词： Plasticity

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Fuel assembly design for supercritical water-cooled reactor

Fuel assembly design for supercritical water-cooled reactor

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作者： Linna, Feng Fawen, Zhu Science and Technology on Reactor System Design Technology Laboratory Third Section of Huafu Road Huayang Town Shuangliu County Chengdu Sichuan Province610213 China

The supercritical water-cooled reactor (SWCR) has been selected as one of the most promising reactors for Generation IV nuclear reactors due to its higher thermal efficiency and more simplified structure compared to the state-of-the-art light water reactors (LWRs). However, there are a large number of potential problems that must be addressed, particularly the fuel assembly design of the SCWR. SCWRs are a kind of high-temperature, high-pressure, water-cooled reactor that operates above the thermodynamic critical point of water (374°C, 22.1 MPa). Corrosion and degradation of materials used in supercritical water environments are determined by several environment- and material-dependent factors. In particular, irradiation-induced changes in microstructure and microchemistry are major concerns in a nuclear reactor. Many structural materials including alloys and ceramics have been proposed for use as SCWR components or materials for applying protective coatings in SCWRs. In this paper, the present status of supercritical fuel assembly design at home and abroad is reported. According to the special requirements of supercritical core design, a kind of configuration design of fuel assembly with two-flow core and using SiC as cladding material are proposed. The analysis results have shown that the design basically meets the requirements of fuel assembly design, which has good feasibility and performance. Copyright © 2015 by ASME.

关键词： Silicon carbide

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Modification of severe accident analysis code for accident tolerant fuel (ATF) applications 27

Modification of severe accident analysis code for accident t...

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

作者： Xu, Tao Ma, Haifu Tian, Chao Wu, Yirui Yang, Junyun Wang, Kai Zhang, Bin Wang, Junlong Deng, Yaxin Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu Sichuan China School of Nuclear Science and Technology Xi'an Jiaotong University Xi'an Shaanxi China

ISBN: (纸本)9784888983051

After the Fukushima nuclear accident, we gradually realized the safety issues of the traditional UO2+Zr nuclear fuel system. Accident tolerance fuel (ATF) is a new generation fuel concept proposed to improve the performance of fuel elements against severe accidents, and can withstand the consequences of accidents for a long time while maintaining or improving their performance under normal operating conditions. Nowadays, the issue of the safe utilization of nuclear energy has attracted more and more attention. reactor severe accident simulation and analysis for ATF has also become one of the important research directions. MELCOR is a fully integrated engineering-grade computer code which can model severe accidents phenomena in light water reactors. However, the existing MELCOR code is only applicable for the UO2+Zr fuel system, and cannot simulate the behavior of the ATF. So it is necessary to modify the MELCOR code for ATF applications. Based on extensive investigation, multi-fuel and multi-cladding processing module and new physical properties of ATF are added to MELCOR. Also the heat transfer model, oxidation model and cladding failure model are modified, enabling the MELCOR code to calculate ATF behaviors. Afterwards, the model verification is carried out. Then the traditional UO2+Zr and five ATF combinations are applied to the traditional three-loop PWR nuclear power plant to perform the safety analysis under large break loss of coolant accident (LBLOCA). The calculation results show that ATFs have better accident tolerance ability than UO2+Zr fuel system used in the commercial PWR, and the temperature of the cladding rises more slowly. The cladding failure is also delayed to different extent. The ATF fuel system has a better performance under the severe accident condition. © 2018-2019 by JSME

关键词： Uranium dioxide

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Numerical study on single Droplet's rise in steam separator by using VOF method

Numerical study on single Droplet's rise in steam separator ...

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2014 ANS Winter Meeting and Nuclear technology Expo

作者： Zhang, Di Luo, Qi Huang, Wei Wang, Kan Department of Engineering Physics Tsinghua University Beijing China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China

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Development and Verification of a New Depletion, Activation and Radiation Source Term Calculation Code 23rd

Development and Verification of a New Depletion, Activation ...

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23rd Pacific Basin Nuclear Conference, PBNC 2022

作者： Wen, Xingjian Zhai, Zian Tang, Songqian Tian, Chao Liu, Zhouyu Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Sichuan Chengdu China Xi’an Jiaotong University Xi’anShaanxi China

ISBN: (纸本)9789819910229

Existing depletion and source term calculation codes lack flexible interfaces, which is difficult to meet the actual engineering design needs. Based on the Chebyshev rational approximation method, a new depletion, activation and source term calculation code TIST is developed and verified. Full-fidelity depletion library based on ENDF-VII.0 and EAF2010 contains 3837 isotopes is generated. Flexible interfaces are built-in to support cross sections from different sources for depletion and activation calculations. The EAF2010 library and the cross sections from the high-fidelity code NECP-X are adopted by TIST for the verification in this work. Based on the nuclide inventory of the depletion and activation calculation, the radiation source term calculation capability of TIST is developed, including the radioactivity, the decay heat, the neutron source and the photon source. The comparison results with NECP-X and FISPACT demonstrate good accuracy of TIST. © 2023, The Author(s).

关键词： Chemical activation

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Safety design and evaluation of loss of forced flow accident of CRS1000

Safety design and evaluation of loss of forced flow accident...

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2018 International Congress on Advances in Nuclear Power Plants, ICAPP 2018

作者： Dan, Zhang Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Changshun Road No.328 ShuangLiu County Chengdu610213 China

ISBN: (纸本)9780894487552

CSR1000 (China supercritical water reactor, 1000MWe) was developed by NPIC, as BWR, the pressure vessel's reactor and directly circulate loop was adopted, however, the coolant will encounter double flow pass in the reactor. The passive engineering safety feature was adopted and the reactor residual heat will be removed by natural circulation of coolant. As above character, loss of feed water or loss of offsite power will cause completely loss of forced flow accident, the flow in first pass of core will encounter flow inversion during this course, these factor make the LOFA(Loss Of forced Flow Accident) become one of the most limiting accident in CSR1000. The LOFA was analyzed by APROS. The result shows, during the short time of LOFA, the passive operation of HFT will mitigate the accident, and during long term, the passive residual heat removal system will function and maintain the core within the safety state. © 2018 American Nuclear Society. All rights reserved.

关键词： Coolants

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