• Nuclear Engineering and Technology
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• v.28 no.3
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• pp.257-264
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• 1996

The conventional Equipment Environment Qualification (EEQ) envelope is developed based on the containment responses during the design basis events. The Safety Depressurization System (SDS) design without In-containment Refueling Water Storage Tank (IRWST) adopted in the Ulchin 3&4 challenges the conventional EEQ envelope during long term Feed and Bleed (F&B) operation due to the direct discharge of high mass and energy into the containment. Therefore, it is necessary to confirm that the containment pressure and temperature history during the long term F&B operation does not violate the conventional EEQ envelope. However, this subject has never been quantitatively assessed before. To investigate the success path of long term F&B operation this paper analyzes the thermal hydraulic response of the containment and Reactor Coolant System (RCS) until the completion of depressurization and cooldown of RCS into Shutdown Cooling System (SCS) entry condition. It is found that the SCS entry condition can be reached within 6 hours without violating the EEQ curve by proper operation of SDS valves, High Pressure Safety Injection (HPSI) pumps and active Containment Heat Removal System (CHRS). The suggested strategy not only demonstrates the feasibility of long term F&B operation but also can be utilized in the preparation of Emergency Procedure Guidelines (EPGs)

• Journal of Korean Society of Water and Wastewater
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• v.24 no.1
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• pp.5-14
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• 2010

A pilot-scale biological aerated filter (BAF) was operated with an anaerobic, anoxic and oxic zone at $23{\pm}1^{\circ}C$. The influent sCOD and total nitrogen concentrations in the feedwater were approximately 250 mg/L and 35 mg N/L, respectively. sCOD removal at optimum hydraulic retention time (HRT) of 3 hours with recirculation rates of 100, 200 and 300% in the column was more than 96%. Total nitrogen removal was consistently above 80% for 4 and 6 hours HRT at 300% recirculation. For 3 hours HRT and 300% recirculation, total nitrogen removal was approximately 79%. Based on fitting results, the kinetic parameter values on nitrification and denitrification show that as recirculation rates increased, the rate of ammonia and nitrate transformation increased. The ammonium loading rates for maximum ammonium removed were 0.15 and 0.19 kg $NH_3$-N/$m^3$-day for 100% and 200% recirculation, respectively. The experimental results demonstrated that the BAF can be operated at an HRT of 3 hours with 200 - 300% recirculation rates with more than 96 % removal of sCOD and ammonium, and at least 75% removal of total nitrogen.

• Nuclear Engineering and Technology
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• v.40 no.1
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• pp.37-48
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• 2008

Flow fields inside feeder pipes have been simulated numerically using a CFD (computational fluid dynamics) code to calculate the shear stress distribution, which is the most important factor in predicting the local regions of feeder pipes highly susceptible to FAC (flow-accelerated corrosion)-induced wall thinning. The CFD approach, with schemes used in this study, to simulate the flow situations inside the CANDU feeder pipes has been verified as it showed a good agreement between the investigation results for the failed feedwater pipe at Surry unit 2 plant in the U.S. and the CFD calculation. Sensitivity studies of the three geometrical parameters, such as angle of the first and second bends, length of the first span between the grayloc hub and the first bend, and length of the second span between the first and the second bends have been performed. CFD analysis reveals that the local regions of feeder pipes of Wolsung unit 1 in Korea, on which wall thickness measurements have been performed so far, are not coincident with the worst regions predicted by the present CFD analysis located in the connection region of straight and bend pipe near the inlet part of the bend intrados. Finally, based on the results of the present CFD analysis, a guide to the selection of the weakest local positions where the measurement of wall thickness should be performed with higher priority has been provided.

• Progress in Superconductivity and Cryogenics
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• v.21 no.2
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• pp.22-25
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• 2019

The superconducting magnetic separation system has been developing to separate the iron oxide scale from the feed water of the thermal power plant. The accumulation in the boiler lowers the heat exchange rate or in the worst case damages it. For this reason, in order to prevent scale generation, controlling pH and redox potential is employed. However, these methods are not sufficient and then the chemical cleaning is performed regularly. A superconducting magnetic separation system is investigated for removing iron oxide scale in a feed water system. Water supply conditions of the thermal power plant are as follows, flow rate 400 t / h, flow speed 0.2 m / s, pressure 2 MPa, temperature $160-200^{\circ}C$, amount of scale generation 50 - 120 t / 2 years. The main iron oxide scale is magnetite (ferromagnetic substance) and its particle size is several tens ${\mu}m$. As the first step we are considering to introduce the system to the chemical cleaning process of the thermal power plant instead of the thermal power plant itself. The current status of development will be reported.

• Nuclear Engineering and Technology
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• v.51 no.3
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• pp.676-682
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• 2019

After the Fukushima Daiichi accident, there has been an increasing preference for passive safety features in the nuclear power industry. Some passive safety systems require limited active components to trigger subsequent passive operation. Under very serious accident conditions, passive safety features could be rendered inoperable or damaged. This study evaluates (i) the performance and effectiveness of the passive safety features of iPOWER (innovative Power Reactor), and (ii) whether a severe accident condition could be reached if the passive safety systems are damaged, namely the case of heat exchanger tube rupture. Analysis results show that the reactor coolant system remains in the hot shutdown condition without operator actions or electricity for over 72 h when the passive auxiliary feedwater systems (PAFSs) are operable without damage. However, heat exchanger tube rupture in the PAFS leads to core damage after about 18 h. Such results demonstrate that, to enhance the safety of iPOWER, maintaining the integrity of the PAFS is critical, and therefore additional protections for PAFS are necessary. To improve the reliability of iPOWER, additional battery sets are necessary for the passive safety systems using limited active components for accident mitigation under such extreme circumstances.

• Corrosion Science and Technology
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• v.18 no.4
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• pp.138-147
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• 2019

Pipes and components of the secondary system in the pressurized water reactor (PWR) are mainly comprised of manufactured carbon steel. Thus, the generated carbon steel corrosion products are transported into the steam generator and deposited, thereby deteriorating the integrity of the steam generator. Environmental condition in the secondary system of the PWRs differs across different locations. So, the corrosion rate and types of corrosion products depend on specific locations in the secondary system. In this study, the quantity and chemical compositions of corrosion products generated in various locations that vary in different temperatures and chemistry conditions were investigated. As a result of evaluating the PWR "Unit A" that is in current operation, the amount of corrosion products generated in the section of high temperature feedwater system was identified as the largest source in the secondary system. Major components of corrosion products were iron oxides such as magnetite, hematite, and lepidocrocite.

• Nuclear Engineering and Technology
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• v.51 no.1
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• pp.116-124
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• 2019

$M^2LFR-1000$ is a medium-power modular lead-cooled fast reactor, developed by University of Science and Technology of China (USTC), aiming at achieving a reactor design fulfilling the Gen IV nuclear system requirements and meanwhile emphasizing the optimum safety and economics. In order to evaluate the safety performance of $M^2LFR-1000$ reactor core, three typical transients are selected from initiating events, which are unprotected transient overpower (UTOP), unprotected loss of offsite power (ULOHS+ULOF) and increase of feedwater flowrate with primary pumps trip (IFW+PLOF). These three transients presented and discussed in this paper are performed with the code Analysis of THermal-hydraulics of LEaks and Transients (ATHLET), which is developed by Gesellschaft $f{\ddot{u}}r$ Anlagen-und Reaktorsicherheit gGmbH (GRS). The results indicate that the $M^2LFR$ is safe enough under these three transients due to the good inherent safety features of the reactor, without human intervention, the reactor will reach a new steady state under UTOP condition.

• Nuclear Engineering and Technology
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• v.50 no.7
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• pp.1173-1183
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• 2018

Transient fluid velocity and pressure fields in a pressurized water reactor (PWR) steam generator (SG) secondary side during the blowdown period of a feedwater line break (FWLB) accident were numerically simulated employing the saturated water flashing model. This model is based on the assumption that compressed water in the SG is saturated at the beginning and decompresses into the two-phase region where saturated vapor forms, creating a mixture of steam bubbles in water by bulk boiling. The numerical calculations were performed for two cases of which the outflow boundary conditions are different from each other; one is specified as the direct blowdown discharge to the atmosphere and the other is specified as the blowdown discharge to an extended calculation domain with atmospheric pressure on its boundary. The present simulation results obtained using the two different outflow boundary conditions were discussed through a comparison with the predictions using a simple non-flashing model neglecting the effects of phase change. In addition, the applicability of each of the non-flashing water discharge and saturated water flashing models for the confirmatory assessments of new SG designs was examined.

• Progress in Superconductivity and Cryogenics
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• v.23 no.3
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• pp.14-19
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• 2021

The reduction of carbon dioxide emissions becomes a global issue, the main source of carbon dioxide emissions in the Asian region is the energy conversion sector, especially coal-fired power plants. We are working to develop technologies that will at least limit the increase in carbon dioxide emissions from the thermal power plants as one way to reduce carbon dioxide emissions. Our research aims to reduce carbon dioxide emissions by removing iron oxide scale from the feedwater system of thermal power plants using a superconducting high-gradient magnetic separation (HGMS) system, thereby reducing the loss of power generation efficiency. In this paper, the background of thermal power plants in Asia is outlined, followed by a case study of the introduction of a chemical cleaning line at an actual thermal power plant in Japan, and the possibility of introducing it into the thermal power plants in China based on the results.

• Journal of the Korean Society of Industry Convergence
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• v.25 no.4_2
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• pp.541-547
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• 2022

Before making a prototype, we predicted the inlet/outlet differential pressure and flow coefficient, which are the most basic design data for the valve through the design and numerical analysis of the trim, which is the most important in the localization development of the 1500Ib high differential pressure control valve used for boiler feed water. As a result, the design value and the analysis value were found to be about 98% similar. The flow field within the fluid velocity of 23m/s to prevent cavitation was also found. The result of the numerical analysis on thermal stress due to the characteristics of valves exposed to high temperatures showed that it was found to be about 18% less than the allowable stress of the bolt fixing the trim. When all loads such as pressure, self-weight, and vibration are applied, however, it is judged to go beyond the currently calculated thermal stress, exceeding the allowable stress.