• Proceedings of the KIEE Conference
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• 2005.10c
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• pp.86-88
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• 2005

The line-start permanent magnet synchronous motor has a high efficiency and an advantage in constant speed operation regardless of the effect of load variation. However it is difficult to predict the performance of characteristics accurately, because of the unbalanced starting torque with the initial starting position of the rotor and the generation of a break torque. In this paper the dynamic characteristics of the line-start permanent magnet synchronous motor arc described and compared with those of the squirrel-cage induction motor through the simulation to find the characteristics of the permanent magnets and the rotor bars in the line-start permanent magnet synchronous motor.

• Proceedings of the Korean Nuclear Society Conference
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• 2002.05a
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• pp.148.2-148
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• 2002

• Transactions of the Korean Society of Mechanical Engineers A
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• v.26 no.4
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• pp.677-682
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• 2002

Break-in is an intentional treatment to enhance the performance life of machinery parts and to maintain static friction behavior. Most studies on break-in have concerned only about surface conditions such as roughness or film formation. But the exact mechanism of break-in has not been found yet. Friction, scuffing behavior and wear of AISI 1045 were studied in relation to break-in and residual stress. The cylinder-on-disk type tribometer was used with the line-contact geometry. Scuffing tests were carried out using a constant load of 730N. In the break-in procedure the step load was applied from 100N to 200N. In this experiment, it was found that the break-in helps compressive residual stress to be formed well enough to enhance the scuffing life during the scuffing test. Specimens that had high compressive residual stress induced by shot-peening show better wear resistance than those were not shot-peened. Results of scuffing test, break-in procedure and wear amount in relation to residual stress have been discussed.

• Journal of computational fluids engineering
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• v.20 no.3
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• pp.87-93
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• 2015

In this study, a numerical analysis was performed to simulate the thermal-hydraulic response of the secondary side of a steam generator(SG) model equipped with an orifice-type SG outlet flow restrictor to a main steam line break(MSLB) at a pressurized water reactor(PWR) plant. The SG analysis model includes the SG upper steam space and the part of the main steam pipe between the SG outlet and the broken pipe end. By comparing the numerical calculation results for the present SG model to those obtained for a simple SG model having no flow restrictor, the effects of the flow restrictor on the thermal-hydraulic response of SG to the MSLB were investigated.

• Journal of Energy Engineering
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• v.23 no.4
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• pp.112-122
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• 2014

A main steam line break (MSLB) test at the ATLAS facility was simulated using the best-estimate thermal-hydraulic system code, MARS-KS. This has been performed as an activity at the third domestic standard problem for code benchmark (DSP-03) that has been organized by Korea Atomic Energy Research Institute (KAERI). The results of the MSLB experiment and the MARS input data prepared for the previous DSP-02 using the ATLAS facility were provided to participants. The preliminary MSLB simulation using the base input data, however, showed unphysical results in the primary-to-secondary heat transfer. To resolve the problems, some improvements were implemented in the MARS input modelling. These include the use of fine meshes for the bottom region of the steam generator secondary side and proper thermal-hydraulics calculation options. Other input model improvements in the heat loss and the flow restrictor models were also made and the results were investigated in detail. From the results of simulations, the limitations and further improvement areas of the MARS code were identified.

• Nuclear Engineering and Technology
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• v.32 no.3
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• pp.214-226
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• 2000

In an effort to assess the performance of KAERI's coupled 3D kinetics - system T/H code, MARS/MASTER, Exercise III of the OECD main steam line break benchmark is solved. The analysis model of the reference plant, TMI-1 - a 2772 MWth B&W plant, consists of three major components: a core neutronics model involving 241$\times$28 neutronic nodes, a vessel 3D T/H model consisting of 374 hydrodynamic volumes, and a 1D system T/H model containing 157 hydrodynamic volumes. The results show that there is a significant amount of flow mixing occurring in the upper and lower plenum regions and the core power distribution evolves to a highly localized shape due to the presence of a stuck rod, as well as the asymmetric flow distribution. It is judged that MARS/MASTER properly captures these drastic 3-dimensional effects. Comparisons with other results submitted to OECD confirm the accuracy of the MARS/MASTER solution.

• Nuclear Engineering and Technology
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• v.53 no.8
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• pp.2477-2487
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• 2021

To improve safety analysis technology for a nuclear reactor containment considering an interaction between a reactor coolant system (RCS) and containment, this study aims at an experimental investigation on the integrated simulation of the RCS and containment, with an integral effect test facility, ATLAS-CUBE. For a realistic simulation of a pressure and temperature (P/T) transient, the containment simulation vessel was designed to preserve a volumetric scale equivalently to the RCS volume scale of ATLAS. Three test cases for a steam line break (SLB) transient were conducted with variation of the initial condition of the passive heat sink or the steam flow direction. The test results indicated a stratified behavior of the steam-gas mixture in the containment following a high-temperature steam injection in prior to the spray injection. The test case with a reduced heat transfer on the passive heat sink showed a faster increase of the P/T inside the containment. The effect of the steam flow direction was also investigated with respect to a multi-dimensional distribution of the local heat transfer on the passive heat sink. The integral effect test data obtained in this study will contribute to validating the evaluation methodology for mass and energy (M/E) and P/T transient of the containment.

• Nuclear Engineering and Technology
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• v.51 no.2
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• pp.402-409
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• 2019

When a loss of coolant accident (LOCA) occurs in a nuclear power plant, accident scenarios which can prevent core damage are defined based on break size. Current probabilistic safety assessment evaluates that core damage can be prevented under small-break LOCA (SBLOCA) and steam generator tube rupture (SGTR) with rapid cool down (RCD) strategy when all safety injection systems are unavailable. However, previous research has pointed out a limitation of RCD in terms of initiation time. Therefore, RCD success criteria estimation based on allowable coping time under a SBLOCA or SGTR when all safety injection systems are unavailable was performed based on time-line and thermal-hydraulic analyses. The time line analysis assumed a single emergency operating procedure flow, and the thermal hydraulic analysis utilized MARS-KS code with variables of break size, cooling rate, and operator allowable time. Results show while RCD is possible under SGTR, it is impossible under SBLOCA at the APR1400's current cooling rate limitation of 55 K/hr. A success criteria map for RCD under SBLOCA is suggested without cooling rate limitation.

• Journal of Energy Engineering
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• v.16 no.3
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• pp.103-112
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• 2007

Flood issue for nuclear power plants designed and built in 1970 is extremely severe for main steam header compartment and main feedwater line region of intermediate building and lower floor. A calculation for flood level at the main feedwater line isolation compartment is now performing by hand calculation. But, this methodology is quite conservative assumption. The goal of this study was to develop method to analyze flowrate using the RETRAN-3D computer code, and the developed method was applied to flood level analysis following main feedwater line break. As a result of analysis, flood level was low remarkably.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05b
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• pp.435-440
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• 1996

Large Loss-of-Coolant Accidents analyzed in Final Safety Analysis Reports are usually covered by Reactor Inlet Header. Reactor Outlet Header and Primary Pump Suction breaks as representative cases. In this study we analyze the total (guillotine) break of an Emergency Core Cooling System (ECCS) pipe located at the ECCS injection point into the Primary Heat Transport System (PHTS). It was expected that thermal-hydraulic behaviors in the PHT and ECC systems are different from those of a Reactor Inlet Header break, having an equivalent break size. The main purpose of this study is to get insights on the differences occurred between the two cases and to assess these differences from the phenomenon behavior point of view. It was also investigated whether the ECCS line break analysis results could be covered by header break analysis results. The study reveals that as the intact loop has almost the same behavior in both analyzed cases. broken loop behavior is different mostly regarding sheath temperature in the critical core pass and pressure decrease in the broken Reactor Inlet Header. Differences are also met in the ECCS behavior and in event sequences timings.