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- 저자 : Xiaoming Chai (검색결과 3 건)
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Jingang Liang,Kan Wang,Yishu Qiu,Xiaoming Chai,Shenglong Qiang 한국원자력학회 2016 Nuclear Engineering and Technology Vol.48 No.3
Because of prohibitive data storage requirements in large-scale simulations, the memoryproblem is an obstacle for Monte Carlo (MC) codes in accomplishing pin-wise threedimensional(3D) full-core calculations, particularly for whole-core depletion analyses. Various kinds of data are evaluated and quantificational total memory requirements areanalyzed based on the Reactor Monte Carlo (RMC) code, showing that tally data, materialdata, and isotope densities in depletion are three major parts of memory storage. Thedomain decomposition method is investigated as a means of saving memory, by dividingspatial geometry into domains that are simulated separately by parallel processors. For thevalidity of particle tracking during transport simulations, particles need to be communicatedbetween domains. In consideration of efficiency, an asynchronous particle communicationalgorithm is designed and implemented. Furthermore, we couple the domain decompositionmethod with MC burnup process, under a strategy of utilizing consistent domain partition inboth transport and depletion modules. A numerical test of 3D full-core burnup calculationsis carried out, indicating that the RMC code, with the domain decomposition method, iscapable of pin-wise full-core burnup calculations with millions of depletion regions.
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Yugao Ma,Jiusong Liu,Hongxing Yu,Changqing Tian,Shanfang Huang,Jian Deng,Xiaoming Chai,Yu Liu,Xiaoqiang He 한국원자력학회 2022 Nuclear Engineering and Technology Vol.54 No.6
The solid-state core of a heat pipe cooled reactor operates at high temperatures over 1000 K withthermal and irradiation-induced expansion during burnup. The expansion changes the gap thicknessbetween the solid components and the material properties, and may even cause the gap closure, whichthen significantly influences the thermal and mechanical characteristics of the reactor core. This studydeveloped an irradiation behavior model for HPRTRAN, a heat pipe reactor system analysis code, tointroduce the irradiation effects such as swelling and creep. The megawatt heat pipe reactor MegaPowerwas chosen as an application case. The coupled irradiation-thermal-mechanical model was developed tosimulate the irradiation effects on the heat transfer and stresses of the whole reactor core. The resultsshow that the irradiation deformation effect is significant, with the irradiation-induced strains up to2.82% for fuel and 0.30% for monolith at the end of the reactor lifetime. The peak temperatures during thelifetime are 1027:3 K for the fuel and 956:2 K for monolith. The gap closure enhances the heat transferbut caused high stresses exceeding the yield strength in the monolith
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Development and validation of reactor nuclear design code CORCA-3D
Ping An,Yongqiang Ma,Peng Xiao,Fengchen Guo,Wei Lu,Xiaoming Chai 한국원자력학회 2019 Nuclear Engineering and Technology Vol.51 No.7
The advanced node core code CORCA-3D is one of the independent developed codes of NPIC for the nuclear reactor core design. CORCA-3D code can calculate the few-group cross section, solve the 3D diffusion equations, consider the thermal-hydraulic feedback, reconstruct the pin-by-pin power. It has lots of functions such as changing core status calculation, critical searching, control rod value calculation, coefficient calculation and so on. The main theory and functions of CORCA-3D code are introduced and validated with a lot of reactor measured data and the SCIENCE system. Now, CORCA-3D code has been applied in ACP type reactor nuclear cores design.
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