Search Results - "Xianan Du"

A new surrogate method for the neutron kinetics calculation of nuclear reactor core transients by Li, Xiaoqi, Zheng, Youqi, Du, Xianan, Xiao, Bowen

“…Reactor core transient calculation is very important for the reactor safety analysis, in which the kernel is neutron kinetics calculation by simulating the…”
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Adaptive energy group division in the few-group cross-section generation for full spectrum reactor modeling with deterministic method by Yang, Yichen, Zheng, Youqi, Du, Xianan, Wu, Hongchun

“…Advanced nuclear reactors, especially the newly developed small and micro-reactors have complex neutron spectrum, which makes the deterministic reactor core…”
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Reactivity Effect Evaluation of Fast Reactor Based on Angular-Dependent Few-Group Cross Sections Generation by Du, Xianan, Wu, Xuewen, Zheng, Youqi, Wang, Yongping

“…Among all the possible occurring reactivity effects of a fast reactor, the situations whereby the control rod was inserted, or the coolant was voided could…”
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The applicability study and validation of TULIP code for full energy range spectrum by Chen, Wenjie, Du, Xianan, Wang, Rong, Zheng, Youqi, Wang, Yongping, Wu, Hongchun

“…NECP-SARAX is a neutronics analysis code system for advanced reactor developed by Nuclear Engineering Computational Physics Laboratory of Xi'an Jiaotong…”
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Verification of a two-step code system MCS/RAST-F to fast reactor core analysis by Tran, Tuan Quoc, Cherezov, Alexey, Du, Xianan, Lee, Deokjung

“…RAST-F is a new full-core analysis code based on the two-step approach that couples a multi-group cross-section generation Monte-Carlo code MCS and a…”
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Conceptual design of a MW heat pipe reactor by Wu, Yunqin, Zheng, Youqi, Chen, Qichang, Li, Jinming, Du, Xianan, Wang, Yongping, Tao, Yushan

“…–In recent years, unmanned underwater vehicles (UUV) have been vigorously developed, and with the continuous deepening of marine exploration, traditional…”
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A new surrogate method for the neutron kinetics calculation of nuclear reactor core transients by Xiaoqi Li, Youqi Zheng, Xianan Du, Bowen Xiao

“…Reactor core transient calculation is very important for the reactor safety analysis, in which the kernel is neutron kinetics calculation by simulating the…”
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Adaptive energy group division in the few-group cross-section generation for full spectrum reactor modeling with deterministic method by Yichen Yang, Youqi Zheng, Xianan Du, Hongchun Wu

“…Advanced nuclear reactors, especially the newly developed small and micro-reactors have complex neutron spectrum, which makes the deterministic reactor core…”
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DEVELOPMENT OF DECAY HEAT MODEL FOR RAST-K by Dzianisau, Siarhei, Park, Jinsu, Choi, Sooyoung, Cherezov, Alexey, Du, Xianan, Lee, Deokjung

“…Decay heat (DH) is the heat produced through a radioactive decay of fission products during or after a reactor operation. It is known as the second largest…”
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Conceptual design of a MW heat pipe reactor by Yunqin Wu, Youqi Zheng, Qichang Chen, Jinming Li, Xianan Du, Yongping Wang, Yushan Tao

“…-In recent years, unmanned underwater vehicles (UUV) have been vigorously developed, and with the continuous deepening of marine exploration, traditional…”
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MACROSCOPIC CROSS SECTIONS GENERATION BY MONTE CARLO CODE MCS FOR FAST REACTOR ANALYSIS by Nguyen, Tung Dong Cao, Lee, Hyunsuk, Du, Xianan, Dos, Vutheam, Tran, Tuan Quoc, Lee, Deokjung

“…Recent researches have become more interested in the feasibility of using Monte Carlo (MC) code to generate multi-group (MG) cross sections (XSs) for fast…”
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The applicability study and validation of TULIP code for full energy range spectrum by Wenjie Chen, Xianan Du, Rong Wang, Youqi Zheng, Yongping Wang, Hongchun Wu

“…NECP-SARAX is a neutronics analysis code system for advanced reactor developed by Nuclear Engineering Computational Physics Laboratory of Xi'an Jiaotong…”
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Application of TULIP/STREAM code in 2-D fast reactor core high-fidelity neutronic analysis by Du, Xianan, Choe, Jiwon, Choi, Sooyoung, Lee, Woonghee, Cherezov, Alexey, Lim, Jaeyong, Lee, Minjae, Lee, Deokjung

“…The deterministic MOC code STREAM of the Computational Reactor Physics and Experiment (CORE) laboratory of Ulsan National Institute of Science and Technology…”
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SARAX: A new code for fast reactor analysis part II: Verification, validation and uncertainty quantification by Zheng, Youqi, Qiao, Liang, Zhai, Zi'an, Du, Xianan, Xu, Zhitao

“…•A new code SARAX has been verified and validated by numerical benchmarks and evaluated benchmark experiments.•The keff bias for different evaluated benchmark…”
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SARAX: A new code for fast reactor analysis part I: Methods by Zheng, Youqi, Du, Xianan, Xu, Zhitao, Zhou, Shencheng, Liu, Yong, Wan, Chenghui, Xu, Longfei

“…•A new code SARAX has been developed for fast reactor neutronics analysis.•A new hybrid method has been proposed to generate the few-group cross section.•The…”
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The acceleration of improved Tone's method in advanced reactor lattice neutron spectrum calculation by Du, Xianan, Yang, Yichen, Zheng, Youqi, Wang, Yongping, Wu, Hongchun

“…This paper focuses on the acceleration of solving fixed-source equations in improved Tone's method. Since each equation is independent of the other with…”
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Development and verification of the coupling code of discrete ordinates and Monte Carlo methods by Zhang, Yunsheng, Wang, Yongping, Du, Xianan, Zheng, Youqi

“…•The discrete ordinates and Monte Carlo coupling method is studied for shielding calculation.•The methods of handling flat, cylindrical, spherical and…”
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Extension of SARAX code system for reactors with intermediate spectrum by Wei, Linfang, Zheng, Youqi, Du, Xianan, Wu, Hongchun

“…•The fast reactor code system SARAX was extended for the intermediate spectrum calculation.•The energy-partitioned resonance self-shielding treatment was used…”
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Multi-parameter nuclear data adjustment for sodium fast reactor by Wang, Yongping, Qiao, Liang, Du, Xianan, Zheng, Youqi

“…This paper studies the application of nuclear data adjustment method in the physics calculation of sodium cooled fast reactor. Based on Bayesian theory, the…”
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The low-enriched uranium core design of a MW heat pipe cooled reactor by Wu, Yunqin, Zheng, Youqi, Tao, Yushan, Liu, Xiaoyue, Du, Xianan, Wang, Yongping

“…•Methods to reduce the fuel enrichment in a heat pipe cooled reactor was researched.•A conceptual design of a heat pipe cooled reactor using low-enriched…”
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