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- 저자 : Du Xianan (검색결과 4 건)
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Application of TULIP/STREAM code in 2-D fast reactor core high-fidelity neutronic analysis
Xianan Du,최지원,최수영,이웅희,Alexey Cherezov,임재용,이민재,이덕중 한국원자력학회 2019 Nuclear Engineering and Technology Vol.51 No.8
The deterministic MOC code STREAM of the Computational Reactor Physics and Experiment (CORE) laboratory of Ulsan National Institute of Science and Technology (UNIST), was initially designed for the calculation of pressurized water reactor two- and three-dimensional assemblies and cores. Since fast reactors play an important role in the generation-IV concept, it was decided that the code should be upgraded for the analysis of fast neutron spectrum reactors. This paper presents a coupled code - TULIP/STREAM, developed for the fast reactor assembly and core calculations. The TULIP code produces self-shielded multi-group cross-sections using a one-dimensional cylindrical model. The generated cross-section library is used in the STREAM code which solves eigenvalue problems for a two-dimensional assembly and a multi-assembly whole reactor core. Multiplication factors and steady-state power dis-tributions were compared with the reference solutions obtained by the continuous energy Monte-Carlo code MCS. With the developed code, a sensitivity study of the number of energy groups, the order of anisotropic PN scattering, and the multi-group cross-section generation model was performed on the keff and power distribution. The 2D core simulation calculations show that the TULIP/STREAM code gives a keff error smaller than 200 pcm and the root mean square errors of the pin-wise power distributions within 2%.
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The applicability study and validation of TULIP code for full energy range spectrum
Chen Wenjie,Du Xianan,Wang Rong,Zheng Youqi,Wang Yongping,Wu Hongchun 한국원자력학회 2023 Nuclear Engineering and Technology Vol.55 No.12
NECP-SARAX is a neutronics analysis code system for advanced reactor developed by Nuclear Engineering Computational Physics Laboratory of Xi’an Jiaotong University. In past few years, improvements have been implemented in TULIP code which is the cross-section generation module of NECP-SARAX, including the treatment of resonance interface, considering the self-shielding effect in non-resonance energy range, hyperfine group method and nuclear library with thermal scattering law. Previous studies show that NECP-SARAX has high performance in both fast and thermal spectrum system analysis. The accuracy of TULIP code in fast and thermal spectrum system analysis is demonstrated preliminarily. However, a systematic verification and validation is still necessary. In order to validate the applicability of TULIP code for full energy range, 147 fast spectrum critical experiment benchmarks and 170 thermal spectrum critical experiment benchmarks were selected from ICSBEP and used for analysis. The keff bias between TULIP code and reference value is less than 300 pcm for all fast spectrum benchmarks. And that bias keeps within 200 pcm for thermal spectrum benchmarks with neutronmoderating materials such as polyethylene, beryllium oxide, etc. The numerical results indicate that TULIP code has good performance for the analysis of fast and thermal spectrum system
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Verification of a two-step code system MCS/RAST-F to fast reactor core analysis
Tuan Quoc Tran,Alexey Cherezov,Xianan Du,이덕중 한국원자력학회 2022 Nuclear Engineering and Technology Vol.54 No.5
RAST-F is a new full-core analysis code based on the two-step approach that couples a multi-group crosssection generation Monte-Carlo code MCS and a multi-group nodal diffusion solver. To demonstrate thefeasibility of using MCS/RAST-F for fast reactor analysis, this paper presents the coupled nodal codeverification results for the MET-1000 and CAR-3600 benchmark cores. Three different multi-group crosssection calculation schemes are employed to improve the agreement between the nodal and referencesolutions. The reference solution is obtained by the MCS code using continuous-energy nuclear data. Additionally, the MCS/RAST-F nodal solution is verified with results based on cross-section generated bycollision probability code TULIP. A good agreement between MCS/RAST-F and reference solution isobserved with less than 120 pcm discrepancy in keff and less than 1.2% root-mean-square error in powerdistribution. This study confirms the two-step approach MCS/RAST-F as a reliable tool for the threedimensional simulation of reactor cores with fast spectrum
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Conceptual design of a MW heat pipe reactor
Wu Yunqin,Zheng Youqi,Chen Qichang,Li Jinming,Du Xianan,Wang Yongping,Tao Yushan 한국원자력학회 2024 Nuclear Engineering and Technology Vol.56 No.3
In recent years, unmanned underwater vehicles (UUV) have been vigorously developed, and with the continuous deepening of marine exploration, traditional energy can no longer meet the energy supply. Nuclear energy can achieve a huge and sustainable energy supply. The heat pipe reactor has no flow system and related auxiliary systems, and the supporting mechanical moving parts are greatly reduced, the noise is relatively small, and the system is simpler and more reliable. It is more favorable for the control of unmanned systems. The use of heat pipe reactors in unmanned underwater vehicles can meet the needs for highly compact, long-life, unmanned, highly reliable, ultra-quiet power supplies. In this paper, a heat pipe reactor scheme named UPR-S that can be applied to unmanned underwater vehicles is designed. The reactor core can provide 1 MW of thermal power, and it can operate at full power for 5 years. UPR-S has negative reactive feedback, it has inherent safety. The temperature and stress of the reactor are within the limits of the material, and the core safety can still be guaranteed when the two heat pipes are failed.
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