• 에너지공학
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• 제12권4호
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• pp.281-288
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• 2003

The MATRA-LMR code has been developed based on a subchannel analysis method for LMR (Liquid Metal Reactor) core subassembly thermal hydraulic design and analysis. The code was improved to allow a seven assembly calculation and can account for inter-assembly heat transfer based on a lumped parameter model. This paper describes the main modifications and improvements of the code and shows reference calculation results which compared single assembly calculation with seven assembly calculation cased for driver and blanket subassemblies of the KALIMER 150 MWe breakeven conceptual design core. KAL- IMER is a pool-type sodium cooled reactor with a thermal output of 392.0 MWth, which have inherently safe, environmentally friendly, proliferation-resistant and economically viable reactor concepts.

• Nuclear Engineering and Technology
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• 제41권5호
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• pp.677-690
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• 2009

A previous study demonstrated that the two-row fuel assembly has much more favorable neutron-physical and thermal-hydraulic behavior than the conventional one-row fuel assemblies. Based on the newly developed two-row fuel assembly, an SCWR core is proposed and analyzed. The performance of the proposed core is investigated with 3-D coupled neutron-physical and thermal-hydraulic calculations. During the coupling procedure, the thermal-hydraulic behavior is analyzed using a sub-channel analysis code and the neutron-physical performance is computed with a 3-D diffusion code. This paper presents the main results achieved thus far related to the distribution of some neutronic and thermal-hydraulic parameters. It shows that with adjustment of the coolant and moderator mass flow in different assemblies, promising neutron-physical and thermal-hydraulic behavior of the SCWR core is achieved. A sensitivity study of the heat transfer correlation is also performed. Since the pin power in fuel assemblies can be non-uniform, a sub-channel analysis is necessary in order to investigate the detailed distribution of thermal-hydraulic parameters in the hottest fuel assembly. The sub-channel analysis is performed based on the bundle averaged parameters obtained with the core analysis. With the sub-channel analysis approach, more precise evaluation of the hot channel factor and maximum cladding surface temperature can be achieved. The difference in the results obtained with both the sub-channel analysis and the fuel assembly homogenized method confirms the importance of the sub-channel analysis.

• Nuclear Engineering and Technology
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• 제54권5호
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• pp.1813-1824
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• 2022

This paper shows the verification work of SARAX code system in the reactor core transient calculation based on the simplified EBR-II Benchmark. The SARAX code system is an analysis package developed by Xi'an Jiaotong University and aims at the advanced reactor R&D. In this work, a neutron-photon coupled power calculation model and a spatial-dependent reactivity feedback model were introduced. To verify the models used in SARAX, the EBR-II SHRT-45R test was simplified to an ULOF transient with an input flowrate change curve by fitting from reference. With the neutron-photon coupled power calculation model, SARAX gave close results in both power fraction and peak power prediction to the reference results. The location of the hottest assembly from SARAX and reference are the same and the relative power deviation of the hottest assembly is 2.6%. As for transient analysis, compared with experimental results and other calculated results, SARAX presents coincident results both in trend and absolute value. The minimum value of core net reactivity during the transient agreed well with the reported results, which ranged from -0.3$ to -0.35$. The results verify the models in SARAX, which are correct and able to simulate the in-core transient with reliable accuracy.

• 방사성폐기물학회지
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• 제21권3호
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• pp.359-370
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• 2023

The thermal integrity of spent nuclear fuels has to be maintained during their long-term dry storage. The detailed temperature distributions of spent fuel assemblies are essential for evaluating the integrity of their dry storage systems. In this study, a subchannel analysis model was developed for a canister of a single fuel assembly using the COBRA-SFS code. The thermal parameters affecting the peak cladding temperature (PCT) of the spent fuel assembly were identified, and sensitivity analyses were performed based on these parameters. The subchannel analysis results indicated the presence of a recirculation flow, based on natural convection, between the fuel assembly and downcomer region. The sensitivity analysis of the thermal parameters indicated that the PCT was affected by the emissivity of the fuel cladding and basket, convective heat transfer coefficient, and thermal conductivity of the fluid. However, the effects of the wall friction factor of the canister, form loss coefficient of the grid spacers, and thermal conductivities of the solid materials, on the PCT were predominantly ignored.

• 한국전산구조공학회:학술대회논문집
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• 한국전산구조공학회 2002년도 가을 학술발표회 논문집
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• pp.131-138
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• 2002

The fuel handling machine handles the core assembly in refueling period of the reactor, it is necessary to predict the motion and structural integrity of it. The kinetic analysis of the fuel handling machine was carried out for the refueling motion. The reaction forces at the joints of machine were calculated with IDEAS code considering the weight of the machine and the loading force of the core assembly, Also, the structural analysis for the machine modeled by lumped-mass and beam elements was performed by using ANSYS code. The stresses and deformations were calculated for the equivalent static force based on the kinetic analysis and the seismic loads. The calculated displacements and stresses are quite low compared with allowable limits.

• 정보보호학회논문지
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• 제31권4호
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• pp.753-762
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• 2021

웹 어셈블리는 웹 환경에서 수행 가능한 명령어 형식을 일컫는다. 최근 웹 어셈블리의 성능적인 우수함 때문에 다양한 웹 애플리케이션에서 웹 어셈블리가 활용되고 있다. 하지만 본 논문에서는 보안 관점에서 웹 어셈블리의 간접호출 명령어에 대한 보호 기능에 취약한 부분이 있다는 것을 알게 되었고, 이에 이러한 웹 어셈블리에서의 간접호출 명령어의 보호를 위한 코드 계측 기술인 WACFI를 제안한다. 구체적으로 WACFI에서는 소스 코드 분석을 통해 얻은 정보를 활용해 웹 어셈블리 코드를 수정하여 웹 어셈블리의 간접호출 명령어 보호 기능을 강화하였다. 우리의 실험결과에 따르면 WACFI는 단지 약 2.75%의 성능 부하만으로 이러한 보안 기능을 제공하는 것으로 확인되었다.

• 한국안전학회지
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• 제22권6호
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• pp.20-26
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• 2007

The new integrated CAD-CAM systems induce an increasing demand for simulation tools, which are able to simulate industrial part assembly processes by welding, gluing, riveting or bolting(more generally by fastening). Concerning fastened flexible parts, there exist no efficient computational aid on tolerance and methodology available on the field. The first part briefly presents the approach method based on the finite element method for TADA(Tolerance Analysis of Deformable Assemblies). The second part compares the results obtained by simulation using the commercial FEM code with the measurements. The principal elements of dispersion have been identified and studied on an experimental basis in order to test the robustness of the TADA model. This has enabled us to verify the model's possibilities as regards industrial constraints such as the use of incompatible meshes or the use of triangular elements and so on.

• 한국군사과학기술학회지
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• 제11권5호
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• pp.107-115
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• 2008

In this paper, we have proposed a wear growth prediction method on the surface of rubber pad of track assembly installed in high-speed battle tank i.e. the automatic model updating code interfacing with commercial finite element simulation software. Also, simple and resonable geometrical, material finite element model was established to be easily updated based on the empirical threshold value of contact pressure on the contact surface. From the iterative model update and analysis results, we discovered a weak point on rubber pad surface and suggested a new design concept for improving the wear performance of track assembly.

• Nuclear Engineering and Technology
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• 제54권7호
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• pp.2670-2689
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• 2022

STREAM - a lattice transport calculation code with method of characteristics for the purpose of light water reactor analysis - has been developed by the Computational Reactor Physics and Experiment laboratory (CORE) of the Ulsan National Institute of Science and Technology (UNIST). Recently, efforts have been taken to develop a photon module in STREAM to assess photon heating and the influence of gamma photon transport on power distributions, as only neutron transport was considered in previous STREAM versions. A multi-group photon library is produced for STREAM based on the ENDF/B-VII.1 library with the use of the library-processing code NJOY. The developed photon solver for the computation of 2D and 3D distributions of photon flux and energy deposition is based on the method of characteristics like the neutron solver. The photon library and photon module produced and implemented for STREAM are verified on VERA pin and assembly problems by comparison with the Monte Carlo code MCS - also developed at UNIST. A short analysis of the impact of photon transport during depletion and thermal hydraulics feedback is presented for a 2D core also from the VERA benchmark.

• Nuclear Engineering and Technology
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• 제53권5호
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• pp.1540-1555
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• 2021

The plate-type fuel assembly adopted in nuclear research reactor suffers from complicated effect induced by non-uniform irradiation, which might affect its stress conditions, mechanical behavior and thermal-hydraulic performance. A reliable numerical method is of great importance to reveal the complex evolution of mechanical deformation, flow redistribution and temperature field for the plate-type fuel assembly under non-uniform irradiation. This paper is the first part of a two-part study developing the numerical methodology for the thermal-fluid-structure coupling behaviors of plate-type fuel assembly under irradiation. In this paper, the thermal-fluid-structure coupling methodology has been developed for plate-type fuel assembly under non-uniform irradiation condition by exchanging thermal-hydraulic and mechanical deformation parameters between Finite Element Model (FEM) software and Computational Fluid Dynamic (CFD) software with Mesh-based parallel Code Coupling Interface (MpCCI), which has been validated with experimental results. Based on the established methodology, the effects of non-uniform irradiation and fluid were discussed, which demonstrated that the maximum mechanical deformation with irradiation was dozens of times larger than that without irradiation and the hydraulic load on fuel plates due to differential pressure played a dominant role in the mechanical deformation.