Citation: | QI Shaopu, YANG Hongyi, YANG Xiaoyan, WANG Xiaokun, WANG Jin, YE Shangshang. Analysis of French Phenix End-of-life Natural Circulation Test Based on FR-Sdaso[J]. Atomic Energy Science and Technology, 2020, 54(2): 273-280. DOI: 10.7538/yzk.2019.youxian.0120 |
[1] |
IAEA. Benchmark analysis of EBR-Ⅱ shutdown heat removal tests, IAEA-TECDOC-1819[R]. Vienna: IAEA, 2017.
|
[2] |
IAEA. Benchmark analyses on the natural circulation test performed during the phenix end-of-life experiments, IAEA-TECDOC-1703[R]. Vienna: IAEA, 2013.
|
[3] |
TENCHINE D, PIALLA D, FANNING T H, et al. International benchmark on the natural convection test in Phenix reactor[J]. Nuclear Engineering and Design, 2013, 258: 189-198.
|
[4] |
BAVIRE R, TAUVERON N, PERDU F, et al. A first system/CFD coupled simulation of a complete nuclear reactor transient using CATHARE2 and TRIO_U: Preliminary validation on the Phénix reactor natural circulation test[J]. Nuclear Engineering and Design, 2014, 277: 124-137.
|
[5] |
JEONG H Y, HA K S, CHOI C W. Multi-dimensional pool analysis of Phenix end-of-life natural circulation test with MARSLMR code[J]. Annals of Nuclear Energy, 2014, 63: 309-316.
|
[6] |
TENCHINE D, PIALLA D, GAUTH P, et al. Natural convection test in Phenix reactor and associated CATHARE calculation[J]. Nuclear Engineering and Design, 2012, 253: 23-31.
|
[7] |
ZHOU Chong, HUBER K, CHENG Xu. Validation of the modified ATHLET code with the natural convection test of the PHENIX reactor[J]. Annals of Nuclear Energy, 2013, 59: 31-46.
|
[8] |
MOCHIZUKI H, KIKUCHI N, LI S. Computation of natural convection test at Phenix reactor using the NETFLOW++ code[J]. Nuclear Engineering and Design, 2013, 262: 1-11.
|
[9] |
SARA S P W, AMMIRABILE L, KLOOSTERMAN J L, et al. Multi-physics models for design basis accident analysis of sodium fast reactors, Part Ⅰ: Validation of three-dimensional TRACE thermal-hydraulics model using Phenix end-of-life experiments[J]. Nuclear Engineering and Design, 2018, 331: 331-341.
|
[10] |
周翀,Klaus Huber,程旭. ATHLET程序的钠冷快堆应用扩展及其验证[J]. 原子能科学技术,2013,47(11):2053-2058.ZHOU Chong, Klaus Huber, CHENG Xu. Modification and validation of ATHLET code for sodium-cooled fast reactor application[J]. Atomic Energy Science and Technology, 2013, 47(11): 2053-2058(in Chinese).
|
[11] |
杨红义,徐銤. OASIS程序的开发与应用[J]. 核科学与工程,2001,21(4):322-325.YANG Hongyi, XU Mi. Development and application of OASIS code under the CEFR project[J]. Chinese Journal of Nuclear Science and Engineering, 2001, 21(4): 322-325(in Chinese).
|
[12] |
任丽霞. 钠冷快堆系统分析程序实用开发[D]. 北京:中国原子能科学研究院,2003.
|
[13] |
任丽霞,王晋,胡文军. 中国实验快堆失去厂外电后单台主泵停运的一回路瞬态特性分析[J]. 核科学与工程,2016,36(1):35-41.REN Lixia, WANG Jin, HU Wenjun. Study for CEFR primary circuit transient performance in case of one primary pump trip during loss of off-site power[J]. Nuclear Science and Engineering, 2016, 36(1): 35-41(in Chinese).
|
[14] |
王晋,张东辉,胡文军. 池式钠冷快堆系统分析程序开发[J]. 原子能科学技术,2016,50(2):198-203.WANG Jin, ZHANG Donghui, HU Wenjun. Development of pool-type sodium-cooled fast reactor system analysis code[J]. Atomic Energy Science and Technology, 2016, 50(2): 198-203(in Chinese).
|
[15] |
杨晓燕,齐少璞,杨红义,等. 不对称工况对池式钠冷快堆堆芯入口温度的影响[J]. 原子能科学技术,2018,52(11):1977-1983.YANG Xiaoyan, QI Shaopu, YANG Hongyi, et al. Impact of asymmetric plant state on core inlet temperature of pool type sodium-cooled fast reactor[J]. Atomic Energy Science and Technology, 2018, 52(11): 1977-1983(in Chinese).
|
[16] |
WANG Jin, ZHANG Donghui, YU Yang. Research on the application of two order discrete scheme in intermediate heat exchanger heat transfer numerical calculation model[C]∥12th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation and Safety. Qingdao: ANS, 2018.
|
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