Citation: | Liu Yapeng, Zhang Dalin, Chen Yutong, Zhou Lei, Tian Wenxi, Qiu Suizheng, Su Guanghui. Numerical Simulation of the Natural Circulation Test of PHENIX Reactor by ACENA[J]. Nuclear Power Engineering, 2024, 45(5): 121-127. doi: 10.13832/j.jnpe.2024.05.0121 |
[1] |
TENCHINE D. Some thermal hydraulic challenges in sodium cooled fast reactors[J]. Nuclear Engineering and Design, 2010, 240(5): 1195-1217. doi: 10.1016/j.nucengdes.2010.01.006
|
[2] |
ONO A, KAMIDE H, KOBAYASHI J, et al. An experimental study on natural circulation decay heat removal system for a loop type fast reactor[J]. Journal of Nuclear Science and Technology, 2016, 53(9): 1385-1396. doi: 10.1080/00223131.2015.1121844
|
[3] |
WATANABE O, OYAMA K, ENDO J, et al. Development of an evaluation methodology for the natural circulation decay heat removal system in a sodium cooled fast reactor[J]. Journal of Nuclear Science and Technology, 2015, 52(9): 1102-1121. doi: 10.1080/00223131.2014.994049
|
[4] |
WANG S B, ZHANG D L, LIU Y P, et al. An experiment‐based validation of a system code for prediction of passive natural circulation in sodium‐cooled fast reactor[J]. International Journal of Energy Research, 2021, 45(8): 12093-12109. doi: 10.1002/er.6103
|
[5] |
IAEA. Benchmark analyses on the natural circulation test performed during the PHENIX end-of-life experiments: IAEA-TECDOC-1703[R]. Vienna: International Atomic Energy Agency, 2013.
|
[6] |
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. doi: 10.1016/j.nucengdes.2013.02.010
|
[7] |
TENCHINE D, PIALLA D, GAUTHÉ P, et al. Natural convection test in Phenix reactor and associated CATHARE calculation[J]. Nuclear Engineering and Design, 2012, 253: 23-31. doi: 10.1016/j.nucengdes.2012.08.001
|
[8] |
齐少璞,杨红义,杨晓燕,等. 基于FR-Sdaso的法国凤凰堆寿期末自然循环实验分析[J]. 原子能科学技术,2020, 54(2): 273-280. doi: 10.7538/yzk.2019.youxian.0120
|
[9] |
YUE N N, ZHANG D L, CHEN J, et al. The development and validation of the inter-wrapper flow model in sodium-cooled fast reactors[J]. Progress in Nuclear Energy, 2018, 108: 54-65. doi: 10.1016/j.pnucene.2018.05.007
|
[10] |
JEONG H Y, HA K S, CHOI C W. Multi-dimensional pool analysis of Phenix end-of-life natural circulation test with MARS-LMR code[J]. Annals of Nuclear Energy, 2014, 63: 309-316. doi: 10.1016/j.anucene.2013.08.010
|
[11] |
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. doi: 10.1016/j.nucengdes.2013.03.056
|
[12] |
ZHOU C, HUBER K, CHENG X. Validation of the modified ATHLET code with the natural convection test of the PHENIX reactor[J]. Annals of Nuclear Energy, 2013, 59: 31-46. doi: 10.1016/j.anucene.2013.03.035
|
[13] |
WAHNON S S P, 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. doi: 10.1016/j.nucengdes.2018.02.038
|
[14] |
PARTHASARATHY U, SUNDARARAJAN T, BALAJI C, et al. Decay heat removal in pool type fast reactor using passive systems[J]. Nuclear Engineering and Design, 2012, 250: 480-499. doi: 10.1016/j.nucengdes.2012.05.014
|
[15] |
BAVIÈRE 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. doi: 10.1016/j.nucengdes.2014.05.031
|
[16] |
PIALLA D, TENCHINE D, LI S, et al. Overview of the system alone and system/CFD coupled calculations of the PHENIX Natural Circulation Test within the THINS project[J]. Nuclear Engineering and Design, 2015, 290: 78-86. doi: 10.1016/j.nucengdes.2014.12.006
|
[17] |
IAEA. Benchmark analyses on the control rod withdrawal tests performed during the PHÉNIX end-of-life experiments: IAEA-TECDOC-1742[R]. Vienna: International Atomic Energy Agency, 2014.
|
[18] |
CHEN Y T, ZHANG D L, LIANG Y, et al. Preliminary development and validation of ACENA code for heavy liquid metal-gas two phase flow simulation[J]. Annals of Nuclear Energy, 2021, 161: 108452. doi: 10.1016/j.anucene.2021.108452
|
[19] |
林悦,张大林,陈宇彤,等. 熔融铅铋与水相互作用热工水力特性研究[J]. 原子能科学技术,2023, 57(S1): 56-66.
|