• Journal of the Society of Naval Architects of Korea
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• v.46 no.6
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• pp.603-610
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• 2009

Ice load is one of the important design parameters for the construction of icebreaking vessels. In this paper, the design ice load prediction for the icebreaking vessels under normal operating condition in ice-covered sea is discussed. The ice loads under normal operating condition are expected from sea trials in moderate ice conditions. In this sense the extreme ice loads during heavy ramming or accidental collision are not considered. Current study describes the global ice load on the hull of the icebreaking vessels. Available ice load data from full-scale sea trials are collected and analyzed according to various ship-ice interaction parameters including displacement, stem angle, speed of a ship and flexural strength and thickness of sea ice. The ice load prediction formula is compared with the collected full-scale sea trials data and it shows a good agreement.

• Journal of the Society of Naval Architects of Korea
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• v.51 no.6
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• pp.489-494
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• 2014

Estimation of correct ice load under various operating conditions is important during the design and the operation stages of an icebreaker. Normal operating conditions are expected from the official field ice trials and also from general ice transit action. In this paper ice load for the Korean icebreaking research vessel, ARAON, under normal operating condition, is discussed. Published ice load data from full-scale sea trials of six icebreakers were analysed to derive an empirical ice load prediction formula. The IBRV ARAON had sea ice trials during 2010 and 2012 summer season. Strain gauge signal were recorded during her icebreaking voyage and the measured strain data were converted to the equivalent hull stress values. The effect of ARAON's speed in ice and the hull stresses are investigated. By comparing the empirical formula and ice load calculation based von measured data, it is recommended to use the empirical ice load estimation formula for the initial design stage.

• Journal of Drive and Control
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• v.21 no.1
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• pp.22-30
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• 2024

Gearboxes designed for reciprocating motion operating mechanisms operate under conditions where both the load and speed undergo continuous variations. When conducting durability tests on gearboxes designed for such applications, operating the target gearbox under conditions similar to the intended usage is essential. The gearbox must be operated for the required number of cycles to validate its durability under conditions mirroring its intended usage. This study devised an accelerated test method for gearboxes, which reduces operating angles and operational strokes. The reliability of the accelerated test was verified by comparing the stresses imposed on the gears under general and acceleration conditions through multi-body dynamic simulations. The results confirmed that the maximum contact stress levels under normal and accelerated conditions were within a 0.1% error range, indicating a minimal difference in the gear damage rates. However, a difference in the maximum contact stress results between the normal and accelerated conditions was observed when inertial forces acted on the output shaft due to the operational acceleration of the gearbox. Therefore, when conducting this acceleration test, caution should be exercised to ensure that the operational load on the gearbox, which affects inertia, does not significantly deviate from the conditions observed under normal operating conditions.

• Nuclear Engineering and Technology
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• v.37 no.6
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• pp.575-586
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• 2005

In a CANDU (CANada Deuterium Uranium) reactor, fuel channel integrity depends on the coolability of the moderator as an ultimate heat sink under transient conditions such as a loss of coolant accident (LOCA) with coincident loss of emergency core cooling (LOECC), as well as normal operating conditions. This study presents assessments of moderator thermal-hydraulic characteristics in the normal operating conditions and one transient condition for CANDU-6 reactors, using a general purpose three-dimensional computational fluid dynamics code. First, an optimized calculation scheme is obtained by many-sided comparisons of the predicted results with the related experimental data, and by evaluating the fluid flow and temperature distributions. Then, using the optimized scheme, analyses of real CANDU-6 in normal operating conditions and the transition condition have been performed. The present model successfully predicted the experimental results and also reasonably assessed the thermal-hydraulic characteristics of a real CANDU-6 with 380 fuel channels. A flow regime map with major parameters representing the flow pattern inside a calandria vessel has also proposed to be used as operational and/or regulatory guidelines.

• Journal of Electrical Engineering and Technology
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• v.3 no.4
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• pp.499-508
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• 2008

Frequency regulation in off-normal conditions has been an important problem in electric power system design/operation and is becoming much more significant today due to the increasing size, changing structure and complexity of interconnected power systems. Increasing economic pressures for power system efficiency and reliability have led to a requirement for maintaining power system frequency closer to nominal value. This paper presents a decentralized frequency control framework using a modified low-order frequency response model containing a proportional-integral(PI) controller. The proposed framework is suitable for near-normal and emergency operating conditions. An $H_{\infty}$ control technique is applied to achieve optimal PI parameters, and an analytic approach is used to analyse the system frequency response for wide area operating conditions. Time-domain simulations with a multi-area power system example show that the simulated results agree with those predicted analytically.

• Nuclear Engineering and Technology
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• v.32 no.6
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• pp.566-594
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• 2000

In most LMFBR(Liquid Metal Fast Breed Reactor) design, the operating temperature is very high and the time-dependent creep and stress-rupture effects become so important in reactor structural design. Therefore, unlike with conventional PWR, the normal operating conditions can be basically dominant design loading because the hold time at elevated temperature condition is so long and enough to result in severe total creep ratcheting strains during total service lifetime. In this paper, elevated temperature design of the conceptually designed baffle annulus regions of KALIMER(Korea Advanced Liquid MEtal Reactor) reactor internal strictures is carried out for normal operating conditions which have the operating temperature 53$0^{\circ}C$ and the total service lifetime of 30 years. For the elevated temperature design of reactor internal structures, the ASME Code Case N-201-4 is used. Using this code, the time-dependent stress limits, the accumulated total inelastic strain during service lifetime, and the creep-fatigue damages are evaluated with the calculation results by the elastic analysis under conservative assumptions. The application procedures of elevated temperature design of the reactor internal structures using ASME Code Case N-201-4 with the elastic analysis method are described step by step in detail. This paper will be useful guide for actual application of elevated temperature design of various reactor types accounting for creep and stress-rupture effects.

• Journal of Korean Society for Atmospheric Environment
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• v.31 no.5
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• pp.461-471
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• 2015

Base on emission factors derived from National Institute of Environmental Research, Particulate matter from combined cycle power plants (CCPPs) has been estimated to be a important source of $PM_{10}$. Generally there is no serious emission of particulate matter in CCPPs. because the fuel of them is natural gas. But emission gas after long shut down season has very high dust content. Therefore $PM_{10}$ emission rate is dependent on its operation mode. In this study, particulate dispersion study for new city near CCPPs complex has performed using CALPUFF model for three case. $PM_{10}$ concentration has big difference between normal operation and 2 case start-up condition after long shutdown. In normal operating conditions, daily $0.32{\sim}0.50{\mu}g/m^3$ influence on of the surrounding area. But when 1~2 aerobic high concentration discharged conditions, average concentration is higher about $9.2{\sim}34.1{\mu}g/m^3$ than normal operating conditions.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.37 no.9
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• pp.1159-1168
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• 2013

In pressurized water nuclear reactors (PWRs), the reactor pressure vessel (RPV) upper head contains penetration nozzles that use a control rod drive mechanism (CRDM). The penetration nozzle uses J-groove weld geometry. Recently, the occurrence of cracking in alloy 600 CRDM penetration nozzle has increased. This is attributable to primary water stress corrosion cracking (PWSCC). PWSCC is known to be susceptible to the welding residual stress and operational stress. Generally, the tensile residual stress is the main factor contributing to crack growth. Therefore, this study investigates the effect on weld residual stress through different analysis methods for normal operating conditions using finite element analysis. In addition, this study also considers the effect of repeated normal operating condition cycles on the weld residual stress. Based on the analysis result, this paper presents a normal operating condition analysis method.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.20 no.3
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• pp.357-364
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• 2022

The safety of a KTC-360 transport cask, a large-capacity pressurized heavy-water reactor transport cask that transports CANDU spent nuclear fuel discharged from the reactor after burning in a pressurized heavy-water reactor, must be demonstrated under the normal transport and accident conditions specified under transport cask regulations. To confirm the thermal integrity of this cask under normal transport and accident conditions, high-temperature and fire tests were performed using a one-third slice model of an actual KTC-360 cask. The results revealed that the surface temperature of the cask was 62℃, indicating that such casks must be transported separately. The highest temperature of the CANDU spent nuclear fuel was predicted to be lower than the melting temperature of Zircaloy-4, which was the sheath material used. Therefore, if normal operating conditions are applied, the thermal integrity of a KTC-360 cask can be maintained under normal transport conditions. The fire test revealed that the maximum temperatures of the structural materials, stainless steel, and carbon steel were 446℃ lower than the permitted maximum temperatures, proving the thermal integrity of the cask under fire accident conditions.

• The Transactions of the Korean Institute of Power Electronics
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• v.18 no.5
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• pp.448-457
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• 2013

This paper proposes a control method under stand-alone operating mode for distributed-module type photovoltaic power conditioners. In conventional schemes, there are some problems of a controller saturation in the DC-link (or load) voltage controller when overly-heavy load, light load, and the generated PV power reduction occurs, as well as when a transition occurs from an overly-loaded to normal conditions. To overcome these problems, the proposed controller method switches the main control target from DC-link voltage to the maximum power point, which is closer to the stable operating point when it returns to normal operating conditions. For the analysis, a state-plane trajectory was given and the circuit analysis by PSIM simulation was done. For the verification, a prototype hardware with 110[W] and 50[W] dual photovoltaic modules has been implemented. From the results, it can be seen that PV power tracking is successfully done with the proposed method even under a stand-alone operation mode.