[1] | Cao Yuming, Zhao Gang, Fang Zheng, Li Yuhuan, Zhou Du, Liu Wenshi, Wang Zheng. Study on Calculation Method of Crack Leakage Rate of Valve Sealing Surface[J]. Nuclear Power Engineering, 2025, 46(1): 136-142. doi: 10.13832/j.jnpe.2025.01.0136 |
[2] | Ye Qian, Tan Chao, Xiong Yan, Li Fei, Shan Fuchang. Development and Validation of Non-Inertial Coordinate System Motion Model Based on System-level 3D Thermal Hydraulic Code[J]. Nuclear Power Engineering, 2024, 45(S1): 78-84. doi: 10.13832/j.jnpe.2024.S1.0078 |
[3] | Jiang Lu, Fu Xiaolong, Zhang Liping, Zhang Ying, Yu Mingda, Tian Jun. Research on Calculation Method of Leakage Rate of Graphite Gasket Seal[J]. Nuclear Power Engineering, 2023, 44(1): 141-147. doi: 10.13832/j.jnpe.2023.01.0141 |
[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] | He Feng, Wu Wanjun, Ma Ruoqun, Fang Yonggang, Ai Honglei, Wang Xinjun, Sun Yingxue. Verification and Research of LBB Pipe Crack Leakage Rate Calculation Software PICLES[J]. Nuclear Power Engineering, 2022, 43(6): 168-173. doi: 10.13832/j.jnpe.2022.06.0168 |
[6] | Guo Yingran, Li Jiangkuan, Lin Meng, Yang Yanhua, Huang Tao. Research of Water Packing Numerical Problem in Thermal Hydraulic System Code[J]. Nuclear Power Engineering, 2021, 42(3): 59-63. doi: 10.13832/j.jnpe.2021.03.0059 |
[7] | Huang Yanping, Zeng Xiaokang, Ding Ji. Simulation Model Architecture and Concept Validation for Thermal Hydraulic Characteristics of Two-Phase Fluid Based on Modelica[J]. Nuclear Power Engineering, 2021, 42(1): 1-7. doi: 10.13832/j.jnpe.2021.01.0001 |
[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] | 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. |
[10] | 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 |
[11] | Wang Xi, Shi Xueming. Thermal-Hydraulic Numerical Study of 3D Model Subcritical Energy Blanket[J]. Nuclear Power Engineering, 2018, 39(1): 28-33. doi: 10.13832/j.jnpe.2018.01.0028 |
[12] | 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 |
[13] | Zhang Yin, Liu Caixia, Zhang Li, Zhou Qi, Han Guosheng, Wei Shuang. A Methodology Study for Deep Penetration Shielding Calculations of Research Reactors Based on MCNP Code[J]. Nuclear Power Engineering, 2016, 37(S1): 75-79. doi: 10.13832/j.jnpe.2016.S1.0075 |
[14] | Wu Wanjun, Huang Xuan, Shen Pingchuan. Study on Calculation Method for Pipe Crack Leakage[J]. Nuclear Power Engineering, 2015, 36(S2): 122-126. doi: 10.13832/j.jnpe.2015.S2.0122 |
[15] | Wu Wanjun, Xie Hai, Lan Bin, Huang Xuan, YE Xianhui. Development of Computer Code on Pipe Crack Leakage[J]. Nuclear Power Engineering, 2015, 36(4): 65-68. doi: 10.13832/j.jnpe.2015.04.0065 |
[16] | Wang Lianjie, Zhao Wenbo, Chen Bingde, Yao Dong, Yang Ping. Development of Coupled 3-D Neutronics/Thermal-Hydraulics Code for SCWR Core Transient Analysis[J]. Nuclear Power Engineering, 2014, 35(S2): 186-189. doi: 10.13832/j.jnpe.2014.S2.0186 |
[17] | PENG Qian, YU Hongxing, Simone VANDROUX, Fabien PERDU, LI Songyu, YANG Wen. Analytical Study on Coupling of CATHARE and TRIOU Code for Nuclear Reactor Thermal-Hydraulic Analysis[J]. Nuclear Power Engineering, 2013, 34(S1): 201-205. |
[18] | ZENG Xiao-kang, LI Yong-liang, YAN Xiao, XIAO Ze-jun, HUANG Yan-ping. Application of CFD Methods in Research of SCWR Thermo-Hydraulics[J]. Nuclear Power Engineering, 2013, 34(1): 114-120. |
[19] | ZHOU Xu-hua, CAI Qi, WANG Deng-ying, LI Fu, YI Xiong-ying. Development and Abecedarian Verification of Reactor Core Physic and Thermal-Hydraulic Analysis Software Package DCNMC[J]. Nuclear Power Engineering, 2013, 34(6): 55-60. |
[20] | MA Yong-qiang, CHAI Xiao-ming, WANG Yu-wei, PAN Jun-jie, AN Ping. Development of Coupled Neutronics/Thermal-Hydraulics CASIR Code System for SCWR Core Steady State[J]. Nuclear Power Engineering, 2013, 34(1): 87-91. |