• 한국초전도학회:학술대회논문집
• /
• 한국초전도학회 2007년도 High Temperature Superconductivity Vol.XVII
• /
• pp.21-21
• /
• 2007

• Nuclear Engineering and Technology
• /
• 제53권9호
• /
• pp.2788-2802
• /
• 2021

Multigroup cross section (MG XS) generation by the UNIST in-house Monte Carlo (MC) code MCS for fast reactor analysis using nodal diffusion codes is reported. The feasibility of the approach is quantified for two sodium fast reactors (SFRs) specified in the OECD/NEA SFR benchmark: a 1000 MW_th metal-fueled SFR (MET-1000) and a 3600 MW_th oxide-fueled SFR (MOX-3600). The accuracy of a few-group XSs generated by MCS is verified using another MC code, Serpent 2. The neutronic steady-state whole-core problem is analyzed using MCS/RAST-K with a 24-group XS set. Various core parameters of interest (core k_eff, power profiles, and reactivity feedback coefficients) are obtained using both MCS/RAST-K and MCS. A code-to-code comparison indicates excellent agreement between the nodal diffusion solution and stochastic solution; the error in the core k_eff is less than 110 pcm, the root-mean-square error of the power profiles is within 1.0%, and the error of the reactivity feedback coefficients is within three standard deviations. Furthermore, using the super-homogenization-corrected XSs improves the prediction accuracy of the control rod worth and power profiles with all rods in. Therefore, the results demonstrate that employing the MCS MG XSs for the nodal diffusion code is feasible for high-fidelity analyses of fast reactors.

• Steel and Composite Structures
• /
• 제35권5호
• /
• pp.641-657
• /
• 2020

A unique lightweight string truss deployable bridge assembled by thin-walled fiber reinforced polymer (FRP) and metal profiles was designed for emergency applications. As a new structure, investigations into the static structural performance under the serviceability limit state are desired for examining the structural integrity of the developed bridge when subjected to unsymmetrical loadings characterized by combined torsion and bending. In this study, a full-scale experimental inspection was conducted on a fabricated bridge, and the combined flexural-torsional behavior was examined in terms of displacement and strains. The experimental structure showed favorable strength and rigidity performances to function as deployable bridge under unsymmetrical loading conditions and should be designed in accordance with the stiffness criterion, the same as that under symmetrical loads. In addition, a finite element model (FEM) with a simple modeling process, which considered the multi segments of the FRP members and realistic nodal stiffness of the complex unique hybrid nodal joints, was constructed and compared against experiments, demonstrating good agreement. A FEM-based numerical analysis was thereafter performed to explore the effect of the change in elastic modulus of different FRP elements on the static deformation of the bridge. The results confirmed that the change in elastic modulus of different types of FRP element members caused remarkable differences on the bending and torsional stiffness of the hybrid bridge. The global stiffness of such a unique bridge can be significantly enhanced by redesigning the critical lower string pull bars using designable FRP profiles with high elastic modulus.

• Nuclear Engineering and Technology
• /
• 제55권11호
• /
• pp.4048-4056
• /
• 2023

The feasibility of using the Monte Carlo code MCS to generate multigroup cross sections for nodal diffusion simulations RAST-F of liquid metal fast reactors is investigated in this paper. The performance of the MCS/RAST-F code system is assessed using steady-state simulations of the ANTS-100e core. The results show good agreement between MCS/RAST-F and MCS reference solutions, with a keff difference of less than 77 pcm and root-mean-square differences in radial and axial power of less than 0.5% and 0.25%, respectively. Furthermore, the MCS/RAST-F reactivity feedback coefficients are within three standard deviations of the MCS coefficients. To validate the internal thermal-hydraulic (TH) feedback capability in RAST-F code, the coupled neutronic/TH1D simulation of ANTS-100e is performed using the case matrix obtained from MCS branch calculations. The results are compared to those obtained using the MARS-LBE system code and show good agreement with relative temperature differences in fuel and coolant of less than 0.8%. This study demonstrates that the MCS/RAST-F code system can produce accurate results for core steady-state neutronic calculations and for coupled neutronic/TH simulations.