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[2] | Chen Xi, Wu Qing, Deng Jian, Liu Yu, Ren Chunming, Wang Xiaoyu, Peng Huanhuan. Development and Validation of DNBR On-line Monitoring System for HPR1000 Reactor[J]. Nuclear Power Engineering, 2024, 45(2): 248-253. doi: 10.13832/j.jnpe.2024.02.0248 |
[3] | Wang Bo, Zhao Wenbo, Zhang Hongbo, Zhao Chen, Chen Zhang, Liu Kun, Zhang Lerui, Gong Zhaohu, Zeng Wei, Li Qing. Validation of HPR1000 Core Modeling and Startup Test with Three-dimensional Characteristic Neutronics Calculation Code SHARK[J]. Nuclear Power Engineering, 2024, 45(S2): 42-48. doi: 10.13832/j.jnpe.2024.S2.0042 |
[4] | Zhao Xuebin, Huang Yanping, Zang Jinguang. Research and Development on Thermal Hydraulic and Safety of Supercritical Water-cooled Reactor[J]. Nuclear Power Engineering, 2023, 44(5): 223-231. doi: 10.13832/j.jnpe.2023.05.0223 |
[5] | Zhang Wei, Li Jingsong, Shi Huilie, Qiao Pengrui, Wang Cong, Zhang Tianqing, He Yingzhao. Thermal-Hydraulic Performance Analysis of Horizontal Spiral Tube Steam Generator for European Lead-Cooled Fast Reactor[J]. Nuclear Power Engineering, 2022, 43(3): 38-45. doi: 10.13832/j.jnpe.2022.03.0038 |
[6] | Chen Baowen, Deng Jian, Ling Yufan, Hu Baolong, Wang Tianshi, Zhu Enping, Wang Ting. Analysis of Blockage Accident of Lead-Based Fast Reactor Single-Box Fuel Assembly Based on CFD[J]. Nuclear Power Engineering, 2021, 42(4): 277-281. doi: 10.13832/j.jnpe.2021.04.0277 |
[7] | Ding Xueyou, Chen Zhiqiang, Wen Qinglong, Ruan Shenhui, Qiao Pengrui. Numerical Investigation on Thermal Hydraulics of Helical Coil Tube Once Through Steam Generator for LBE Fast Reactor[J]. Nuclear Power Engineering, 2021, 42(4): 21-26. doi: 10.13832/j.jnpe.2021.04.0021 |
[8] | Li Jiangkuan, Huang Tao, Lin Meng, Wang Xu, Chen Junjie. Study on Calculation Method of Courant Limit in Thermal Hydraulic System Analysis Code[J]. Nuclear Power Engineering, 2021, 42(4): 63-67. doi: 10.13832/j.jnpe.2021.04.0063 |
[9] | Wang Qilong, Ma Ying, Xing Hui, Sun Chuanyi. Study on Countermeasures for Effect of Flow-Induced Vibration Test of Reactor Internals in First Demonstration Project of HPR1000 on Total Construction Period[J]. Nuclear Power Engineering, 2021, 42(3): 108-116. doi: 10.13832/j.jnpe.2021.03.0108 |
[10] | Wang Kun, Dong Xiucheng, Liu Haipeng, Zhang Xin, Yuan Jiangtao. Three-Dimensional Flow Field Calculation of Pressure Vessel in Small Pressurized Water Reactor[J]. Nuclear Power Engineering, 2020, 41(5): 20-23. |
[11] | Zhao Pengcheng Liu Zijing, Yu Tao, Liu Peiqi, Xie Jinsen, Chen Zhenping, . Research on Multi-Physics Coupling Method of Pool Type Fast Reactor Based on CFD[J]. Nuclear Power Engineering, 2020, 41(6): 36-44. |
[12] | Liu Wei, Zhang Yong, Jiang Xiaowei, Zhang Cheng, Zhang Dalin. Development and Verification of Thermal-Hydraulic Transient Analysis Code in Plate-Type Fuel Nuclear Reactor[J]. Nuclear Power Engineering, 2019, 40(5): 18-22. |
[13] | Huang Zongren, Liu Changwen, Lai Jianyong, Li Feng, Wang Xiaoyu, Li Yan, Li Haiying, Leng Guijun. Research and Design on First of a Kind Test of HPR1000[J]. Nuclear Power Engineering, 2019, 40(5): 184-186. |
[14] | Yu Hongxing, Zhou Jinman, Leng Guijun, Deng Jian, Liu Yu, Wu Qing, Liu Wei. General Technology Features of Reactor Core and Safety Systems Design of HPR1000[J]. Nuclear Power Engineering, 2019, 40(1): 1-7. doi: 10.13832/j.jnpe.2019.01.0001 |
[15] | Tang Huapeng, Liu Yu, Chen Xi, Huang Huijian, Shen Caifen, Li Song. Hydraulic Analysis for HPR1000 Reactor with PHYCA Program[J]. Nuclear Power Engineering, 2019, 40(2): 23-26. doi: 10.13832/j.jnpe.2019.02.0023 |
[16] | Wei Shiying, Wang Chenglong, Su Guanghui, Tian Wenxi, Qiu Suizheng. Development of Analysis Code for Pb-Bi Cooled Direct-Contact-Boiling Water Fast Reactor System[J]. Nuclear Power Engineering, 2018, 39(4): 67-70. doi: 10.13832/j.jnpe.2018.04.0067 |
[17] | Tian Xiaoyan, Jiang Xinbiao, Chen Lixin, Li Huaqi, Yang Ning, Zhu Lei, MA Tengyue. Development of Code for Steady-State Thermal-Hydraulic Analysis in Bimodal Space Nuclear Reactor with Heat Pipe[J]. Nuclear Power Engineering, 2017, 38(5): 34-39. doi: 10.13832/j.jnpe.2017.05.0034 |
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[20] | ZHANG Dan, LIU Changwen, LU Jianchao. Study on Heat Transfer and Hydraulic Model of Spiral-Fin Fuel Rods Based on Equivalent Annulus Method[J]. Nuclear Power Engineering, 2011, 32(1): 81-84,94. |