Results 31 to 40 of about 130 (115)

Numerical simulation of decay heat scattering out of AP1000 nuclear reactor spent fuel cask based on finite element method

Thermal Science, 2020
The heat transfer performance of the spent fuel transport cask is inseparably related to the safety of the whole reprocessing system. In this study, we carried out the thermal analysis on the NAC-STC transport cask for AP1000 spent fuel assembly to evaluate the thermal performance of transport cask by the finite element method software ANSYS.
Shujian Tian, Guozhuang Li, Jun Liu, Haoyun Yuan, Xiaoyong Song, Weishu Wang   +5 more
openaire   +2 more sources

Evaluation of Passive Safety Injection System Performance Under Large Break LOCA for Qinshan PWR

Frontiers in Energy Research, 2021
In this work, a brand new passive safety injection system has been designed for the ocean-based Qinshan Phase I nuclear power plant to update and replace the traditional active ones. The passive safety injection system is made up of high pressure, medium
Xuesong Wang, Lin Sun, Meiru Liu, Genglei Xia   +3 more
doaj   +1 more source

The General Design and Technology Innovations of CAP1400

Engineering, 2016
The pressurized water reactor CAP1400 is one of the sixteen National Science and Technology Major Projects. Developed from China's nuclear R&D system and manufacturing capability, as well as AP1000 technology introduction and assimilation, CAP1400 is an ...
Mingguang Zheng, Jinquan Yan, Shentu Jun, Lin Tian, Xujia Wang, Zhongming Qiu   +5 more
doaj   +1 more source

Study on bidirectional fluid-solid coupling characteristics of reactor coolant pump under steady-state condition

Nuclear Engineering and Technology, 2019
The AP1000 reactor coolant pump is a vertical shielded-mixed flow pump, is the most important coolant power supply and energy exchange equipment in nuclear reactor primary circuit system, whose steady-state and transient performance affect the safety of ...
Xiuli Wang, Yonggang Lu, Rongsheng Zhu, Qiang Fu, Haoqian Yu, Yiming Chen   +5 more
doaj   +1 more source

Performance Analysis of AP1000 Passive Systems during Direct Vessel Injection (DVI) Line Break

Atom Indonesia, 2016
Generation II Nuclear Power Plants (NPPs) have a design weakness as shown by the Fukushima accident. Therefore, Generation III+ NPPs are developed with focus on improvements of fuel technology and thermal efficiency, standardized design, and the use of ...
A.S. Ekariansyah, S. Widodo
doaj   +1 more source

Research on rapid source term estimation in nuclear accident emergency decision for pressurized water reactor based on Bayesian network

Nuclear Engineering and Technology, 2021
Nuclear emergency preparedness and response is an essential part to ensure the safety of nuclear power plant (NPP). Key support technologies of nuclear emergency decision-making usually consist of accident diagnosis, source term estimation, accident ...
Guohua Wu, Jiejuan Tong, Liguo Zhang, Diping Yuan, Yiqing Xiao   +4 more
doaj   +1 more source

Methodology for Sensitivity Analysis of Homogenized Cross-Sections to Instantaneous and Historical Lattice Conditions with Application to AP1000^® PWR Lattice

Energies, 2021
In the two-step method for nuclear reactor simulation, lattice physics calculations are performed to compute homogenized cross-sections for a variety of burnups and lattice configurations. A nodal code is then used to perform full-core analysis using the
Dean Price, Thomas Folk, Matthew Duschenes, Krishna Garikipati, Brendan Kochunas   +4 more
doaj   +1 more source

Development and Application of CAP1400 Numerical Reactor System

Yuanzineng kexue jishu, 2022
Numerical reactor technology describes a variety of physical phenomena in the core of a nuclear reactor through highprecision, highresolution, high confidence and high fidelity numerical simulation method, based on highperformance calculation and ...
CAO Liangzhi;DENG Li;YANG Bo;LIU Zhouyu;LIU Peng;TANG Chuntao;SHI Dunfu;CHEN Ronghua;TIAN Wenxi;PENG Lianghui;WAN Chenghui;ZHANG Minwan;BI Guangwen;FEI Jingran;XU Xiaobei;LI Fan
doaj  

Research on the uncertainty problem of SDG fault diagnosis based on information flow

Frontiers in Energy Research, 2023
Accurate and complete diagnosis of nuclear accidents is the primary link to nuclear emergencies. The signed directed graph (SDG), as a common method of fault diagnosis, has good completeness.
Jiahui Huang, Jie Liu, Xiaohua Yang, Jiayu Ma, Hua Liu   +4 more
doaj   +1 more source

Establishing and Assessing of a Leak Collection System Analysis Method by Computational Fluid Dynamics

Modelling and Simulation in Engineering, Volume 2025, Issue 1, 2025.
Small leaks within a nuclear power plant can escalate into significant leaks, leading to plant shutdown and substantial losses if operational limits are exceeded. Thus, the demand for systems that can rapidly detect minor leaks is increasing. Current research seeks to address this need.
Dae Kyung Choi, Won Man Park, Woo-Shik Kim, Dong-Jin Euh, Tae-Soon Kwon, Choengryul Choi, Arpan Hazra   +6 more
wiley   +1 more source