• 한국전산유체공학회:학술대회논문집
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• 2008.03b
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• pp.444-447
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• 2008

A nuclear fuel test loop (after below, FTL) is installed in IR1 of an irradiation hole in HANARO for testing neutron irradiation characteristics and thermo hydraulic characteristics of a fuel loaded in a light water power reactor (PWR) or a heavy water power reactor (CANDU). There is an in-pile section (IPS) and an out-pile section (OPS) in this test loop. When HANARO is normally operated, the fuel loaded in the IPS has a nuclear reaction heat generated by a neutron irradiation. To remove the generated heat and to maintain an operation condition of the test fuel, a main cooling water system (MCWS) is installed in the OPS of the FTL. The pump can not continuously suck a fluid and not pressurize the fluid during a cold function test. To verify the flow characteristics of the MCWS, a flow net work analysis has been conducted. When the higher elevation pipelines wholly filled with coolant, it was confirmed through the analysis results that the pump pressurized the coolant normally. And the analysis results described the system characteristics with operation temperature and pressure variation satisfactorily.

• Korean Journal of Materials Research
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• v.9 no.7
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• pp.715-723
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• 1999

Magnesium has been used as wheel materials in the automotive industry for more than 20 years. The magnesium wheels, which are lighter by 25% than aluminum wheels, provide easy controllability providing excellent road holding by the reduction of weight. The purpose of this work is to develop cast AZ91D alloy wheel by sand cast and permanent mold cast. The fluxless melting with the protective gas $(SF_6+CO_2)$ was Performed to eliminate oxidation of melt and impurity. The transfer of molten magnesium to the mold was done by using gas-pressurized Pump system through the heated pipe. The mechanical properites of AZ91D alloy wheel were investigated as a function of heat treatment, ingot composition.

• Journal of Welding and Joining
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• v.23 no.6
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• pp.61-66
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• 2005

A probabilistic approach for evaluating failure risk is suggested in this paper. Probabilistic fracture analyses were performed for a pressurized pipe of a Cr-Mo steel reflecting variation of material properties at high temperature. A crack was assumed to be located along the weld fusion line. Probability density functions of major variables were determined by statistical analyses of material creep and creep crack growth data measured by the previous experimental studies by authors. Distributions of these variables were implemented in Monte Carlo simulation of this study. As a fracture parameter for characterizing growth of a fusion line crack between two materials with different creep properties, $C_t$ normalized with $C^*$ was employed. And the elapsed time was also normalized with tT, Resultingly, failure probability as a function of operating time was evaluated fur various cases. Conventional deterministic life assessment result was turned out to be conservative compared with that of probabilistic result. Sensitivity analysis for each input variable was conducted to understand the most influencing variable to the analysis results. Internal pressure, creep crack growth coefficient and creep coefficient were more sensitive to failure probability than other variables.

• Proceedings of the Korean Society of Precision Engineering Conference
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• 2005.10a
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• pp.808-813
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• 2005

Structural integrity assessment of defected pipe is important in fitness for service evaluation and proper engineering assessment is needed to determine whether pipelines are still fit for service. This paper present a failure prediction of gas pipes made of APIl X65 steel with gouge using stress-modified true fracture strain, which is regarded as a criterion of ductile fracture. For this purpose, API X65 pipes with gouge are simulated using elastic-plastic FE analyses with the proposed ductile failure criterion and the resulting burst pressures are compared with experimental data. Agreements are quite good, which gives confidence in the use of the proposed criteria to defect assessment fer gas pipelines. Then, further extensive finite element analyses are performed to obtain the burst pressure solution of pipes with gouge as a function of defect depth, length and pipeline geometry.

• Journal of Theoretical and Applied Mechanics
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• v.1 no.1
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• pp.97-109
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• 1995

In this paper, a finite viscoelastic continuum model for rubber and its finite element analysis are presented. This finite viscoelatic model based on continuum mechanics is an extended model of Johnson and Wuigley's 1-D model. In this extended model, continuum based kinematic measures are rigorously defied and by using this kinematic measures, elastic stage law and flow rule are introduced. In kinematics, three configuration are introduced. In kinematics, three configuration are introduced. They are reference, current and virtual visco configurations. In elastic state law, it is assumed that at a certain time, there exists an elastic potential which describes the recoverable elastic energy. From this elastic potential, elastic state law is derived. The proposed flow rule is based on phenomenological observation. The flow rule gives precise relaxation response. In finite element approximation, mixed Lagrangian description is used, where total and similar method of updated Lagrangian descriptions are used together. This approach reduces the numerical job and gives simple nonlinear syatem of equations. To satisfy the incompressible condition, penalty-type modified Mooney-Rivlin energy function is adopted. By this method nearly incompressible condition is obtain the virtual visco configuration. For verification, uniaxial stretch tests are simulated for various stretch rates. The simulated results show good agreement with experiments. As a practical experiments. As a preactical example, pressurized rubber plate is simulated. The result shows finite viscoelastic effects clearly.

• Nuclear Engineering and Technology
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• v.51 no.4
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• pp.1060-1068
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• 2019

This work aims to establish a model of a primary water stress corrosion crack growth rate of Alloy 690 material for the head penetration nozzles of Korean pressurized water reactors. The test material had an inhomogeneous microstructure with bands of fine-grains and intragranular carbides in the matrix of coarse-grains, which was similar to the archive materials of the head penetration nozzles. The crack growth rate was measured from the strain-hardened materials as a function of the stress intensity factor in simulated primary water at various temperatures and dissolved hydrogen contents. The effects of strain-hardening, temperature, and dissolved hydrogen on the crack growth rate were analyzed independently, and were then introduced as normalizing factors in the crack growth rate model. The crack growth rate model proposed in this work provides a key element of the tools needed to assess the progress of a stress corrosion crack when detected in thick-wall Alloy 690 components in Korean reactors.

• Nuclear Engineering and Technology
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• v.55 no.9
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• pp.3213-3228
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• 2023

As an important part of the digital reactor, the pin-by-pin wise fine coupling calculation is a research hotspot in the field of nuclear engineering in recent years. It provides more precise and realistic simulation results for reactor design, operation and safety evaluation. CORCA-K a nodal code is redeveloped as a robust pin-by-pin wise neutronics and thermal-hydraulic coupled calculation code for pressurized water reactor (PWR) core. The nodal green's function method (NGFM) is used to solve the three-dimensional space-time neutron dynamics equation, and the single-phase single channel model and one-dimensional heat conduction model are used to solve the fluid field and fuel temperature field. The mesh scale of reactor core simulation is raised from the nodal-wise to the pin-wise. It is verified by two benchmarks: NEACRP 3D PWR and PWR MOX/UO₂. The results show that: 1) the pin-by-pin wise coupling calculation system has good accuracy and can accurately simulate the key parameters in steady-state and transient coupling conditions, which is in good agreement with the reference results; 2) Compared with the nodal-wise coupling calculation, the pin-by-pin wise coupling calculation improves the fuel peak temperature, the range of power distribution is expanded, and the lower limit is reduced more.

• Nuclear Engineering and Technology
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• v.55 no.10
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• pp.3754-3767
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• 2023

Turbulent jet discharges in subcooled water pools are essential for safety systems in nuclear power plants, specifically in the pressure suppression pool of boiling water reactors and In-containment Refueling Water Storage Tank of advanced pressurized water reactors. The gas and liquid flow in these systems is investigated using multiphase flow analysis. This field has been extensively examined using a combination of experiments, theoretical models, and Computational Fluid Dynamics (CFD) simulations. ANSYS CFX offers two approaches to model multiphase flow behavior. The non-homogeneous Eulerian-Eulerian Model has been used in this work; it computes global information and is more convenient to study interpenetrated fluids. This study utilized the Large Eddy Simulation Model as the turbulence model, as it is better suited for non-stationary and buoyant flows. The CFD results of this study were validated with experimental data and theoretical results previously obtained. The figures of merit dimensionless penetration length and the dimensionless buoyancy length show good agreement with the experimental measurements. Correlations for these variables were obtained as a function of dimensionless numbers to give generality using only initial boundary conditions. CFD numerical model developed in this research has the capability to simulate the behavior of non-condensable gases discharged in water.

• Nuclear Engineering and Technology
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• v.55 no.10
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• pp.3525-3534
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• 2023

Non-uniformity of Pu and Cm composition in used nuclear fuel was analyzed to determine its effect on Pu accountancy in pyroprocessing, while employing the Pu-to-²⁴⁴Cm ratio method. Burnup simulation of a typical pressurized water reactor fuel assembly, required for the analysis, was carried out using MCNP code. Used fuel nuclide composition, as a function of nine axial and two radial meshes, were evaluated. The axial variation of neutron flux and self-shielding effects were found to affect the uniformity of Pu and Cm compositions and in turn the Pu-to-²⁴⁴Cm ratio. However, the results of the study showed that these non-uniformities do not affect the use of Pu-to-²⁴⁴Cm ratio method for Pu accountancy, if the measurement samples are drawn from the voloxidized powder at the feed step of pyroprocessing. 'Material Unaccounted For' and its uncertainty estimates are also presented for a pyrprocessing facility to verify safeguards monitoring requirements of the IAEA.

• Nuclear Engineering and Technology
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• v.56 no.6
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• pp.2375-2387
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• 2024

The passive safety systems (PSSs) within water-cooled reactors are meticulously engineered to function autonomously, requiring no external power source or manual intervention. They depend exclusively on inherent natural forces and the fundamental principles of reactor physics, such as gravity, natural convection, and phase changes, to manage, alleviate, and avert the release of radioactive materials into the environment during accident scenarios like a loss-of-coolant accident (LOCA). PSSs are already integrated into such operating commercial reactors as the Advanced Pressurized Reactor-1000 MWe (AP1000) and the Water-Water Energetic Reactor-1200 MWe (WWER-1200) are adopted in most of the upcoming small modular reactor (SMR) designs. Examples of water-cooled SMR PSSs are the passive emergency core-cooling system (ECCS), passive containment cooling system (PCCS), and passive decay-heat removal system, the designs of which vary based on reactor system-design requirements. However, understanding the accident-event progression and phases of a LOCA is pivotal for adopting a specific PSS for a new SMR design. This study covers the accident-event progression for direct vessel injection (DVI) small-break loss-of-coolant accident (SB-LOCA), associated physics phenomena, knowledge gaps, and important figures of merit (FOMs) that may need to be evaluated and assessed to validate thermal-hydraulics models with an available experimental dataset to support new SMR design and development.