• Journal of the Korean institute of surface engineering
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• v.39 no.1
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• pp.35-42
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• 2006

In this study, the fatigue properties on the cavitation damage of the flame quenched 8.8Al-bronze (8.8Al-4.5Ni-4.5Fe-Cu) as well as the current nuclear pump impeller materials (8.8Al-bronze, STS316 and SR50A) has been investigated using an ultrasonic vibratory cavitation test. For this the impact loads of cavitation bubbles generated by ultrasonic vibratory device quantitatively evaluated and simultaneously the cavitation erosion experiments have been carried out. The fatigue analysis on the cavitation damage of the materials has been made from the determined impact load distribution (e.g. impact load, bubble count) and erosion parameters (e.g. incubation period, MDPR). According to Miner's law, the determined exponents b of the F-N relation ($F^b$ N = Constant) at the incubation stage (N: the number of fracture cycle) were 5.62, 4.16, 6.25 and 8.1 for the 8.8Al-bronze, flame quenched one, STS316 and SR50A alloys. respectively. At the steady state period, the exponents b of the F-N' curve (N': the number of cycle required for $1{\mu}m$ increment of MDP) were determined as 6.32, 5, 7.14 and 7.76 for the 8.8Al-bronze, flame quenched one, STS316, and SR50A alloys, respectively.

• Nuclear Engineering and Technology
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• v.52 no.11
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• pp.2471-2478
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• 2020

Pumps are essential machinery in the various industries. With the development of high-speed and large-scale pumps, especially high energy density, high requirements have been imposed on the vibration and noise performance of pumps, and cavitation is an important source of vibration and noise excitation in pumps, so it is necessary to improve pumps cavitation performance. The modern pump optimization design method mainly adopts parameterization and artificial intelligence coupling optimization, which requires direct correlation between geometric parameters and pump performance. The existing cavitation performance calculation method is difficult to be integrated into multi-objective automatic coupling optimization. Therefore, a fast prediction method for pump cavitation performance is urgently needed. This paper proposes a novel cavitation prediction method based on impeller pressure isosurface at single-phase media. When the cavitation occurs, the area of pressure isosurface S_iso increases linearly with the NPSH_a decrease. This demonstrates that with the development of cavitation, the variation law of the head with the NPSH_a and the variation law of the head with the area of pressure isosurface are consistent. Therefore, the area of pressure isosurface S_iso can be used to predict cavitation performance. For a certain impeller blade, since the area ratio R_s is proportional to the area of pressure isosurface S_iso, the cavitation performance can be predicted by the R_s. In this paper, a new cavitation performance prediction method is proposed, and the feasibility of this method is demonstrated in combination with experiments, which will greatly accelerate the pump hydraulic optimization design.

• Nuclear Engineering and Technology
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• v.54 no.12
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• pp.4571-4584
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• 2022

The analysis of rapid flow transients in Reactor Coolant Pumps (RCP) is essential for a reactor safety study. An accurate and precise analysis of the RCP coastdown is necessary for the reactor design. The coastdown of RCP affects the coolant temperature and the colloidal crud in the primary coolant. A realistic and kinetic model has been used to investigate the behavior of activated colloidal crud in the primary coolant and steam generator that solves the pump speed analytically. The analytic solution of the non-dimensional flow rate has been determined by the energy ratio β. The kinetic energy of the coolant fluid and the kinetic energy stored in the rotating parts of a pump are two essential parameters in the form of β. Under normal operation, the pump's speed and moment of inertia are constant. However, in a coastdown situation, kinetic damping in the interval has been implemented. A dynamic model ACCP-SMART has been developed for System Integrated Modular and Advanced Reactor (SMART) to investigate the corrosion due to activated colloidal crud. The Fickian diffusion model has been implemented as the reference corrosion model for the constituent component of the primary loop of the SMART reactor. The activated colloidal crud activity in the primary coolant and steam generator of the SMART reactor has been studied for different equilibrium corrosion rates, linear increase in corrosion rate, and dynamic RCP coastdown situation energy ratio b. The coolant specific activity of SMART reactor equilibrium corrosion (4.0 mg s^-1) has been found 9.63×10^-3 µCi cm^-3, 3.53×10^-3 µC cm^-3, 2.39×10^-2 µC cm^-3, 8.10×10^-3 µC cm^-3, 6.77× 10^-3 µC cm^-3, 4.95×10^-4 µC cm^-3, 1.19×10^-3 µC cm^-3, and 7.87×10^-4 µC cm^-3 for ²⁴Na, ⁵⁴Mn, ⁵⁶Mn, ⁵⁹Fe, ⁵⁸Co, ⁶⁰Co, ⁹⁹Mo, and ⁵¹Cr which are 14.95%, 5.48%, 37.08%, 12.57%, 10.51%, 0.77%, 18.50%, and 0.12% respectively. For linear and exponential coastdown with a constant corrosion rate, the total coolant and steam generator activity approaches a higher saturation value than the normal values. The coolant and steam generator activity changes considerably with kinetic corrosion rate, equilibrium corrosion, growth of corrosion rate (ΔC/Δt), and RCP coastdown situations. The effect of the RCP coastdown on the specific activity of the steam generators is smeared by linearly rising corrosion rates, equilibrium corrosion, and rapid coasting down of the RCP. However, the time taken to reach the saturation activity is also influenced by the slope of corrosion rate, coastdown situation, equilibrium corrosion rate, and energy ratio β.

• Transactions of the Korean Society for Noise and Vibration Engineering
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• v.21 no.12
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• pp.1098-1103
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• 2011

The structural integrity of APR1400 reactor vessel internals has been being assessed referring the US Nuclear Regulatory Commission regulatory guide 1.20, comprehensive vibration assessment program. The program is composed of a vibration and stress analysis, a vibration and stress measurement, and an inspection. This paper covers the vibration and stress analysis on the reactor vessel internals by the pulsation of reactor coolant pump. 3-dimensional models to calculate the hydraulic loads and structural responses were built and the pressure distributions and the structural responses were predicted using ANSYS. This paper presents that APR1400 reactor vessel internals have enough structural integrity against the pulsation of reactor coolant pump as the peak stress of the reactor vessel internals is much lower than the acceptance limit.

• Journal of Advanced Marine Engineering and Technology
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• v.31 no.1
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• pp.84-94
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• 2007

As a reactor coolant pump(RCP) is operated in the nuclear power system for a long time. so its surface is continuously contaminated by radioactive scales. In order to perform regular or emergency repair about RCP internals a special decontamination process should be used to reduce the radiation from the RCP surface by means of chemical cleaning. In this study, applicable possibility in chemical decontamination for RCP was investigated on the various materials. The STS 304 showed the best electrochemical properties for corrosion resistance than other materials. However, the pitting corrosion was slightly generated in both STS 415 and STS 431 with the increasing numbers of cycle and intergranular corrosion were sporadically observed. The size of their pitting corrosion and intergranular corrosion were also increased with increasing cycle numbers.

• Journal of the Korean Society of Safety
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• v.12 no.4
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• pp.21-26
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• 1997

Korea Atomic Energy Research Institute (KAERI) developed a risk monitor called Risk Monster which supports for plant operators and maintenance schedulers to monitor plant risk and to avoid high peak risk by rearranging maintenance work schedule. Risk Monster can update the plant risk continuously according to the change of system/component configuration since Risk Monster reevaluates the plant risk based on the Probabilistic Safety Assessment (PSA) results. A brief description of Risk Monster is provided. The PSA model of UCN 3, 4 nuclear power plant was converted by KAERI to Risk Monster model. Using this Risk Monster model, a feasibility study of the on-line maintenance of an Essential Service Water (ESW) pump was performed. On-line maintenance of one ESW pump has been shown to be acceptably safe, and has economic benefits. In addition, it is not a violation of technical specification to continue plant operation with an out-of-service ESW pump.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.16 no.3
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• pp.389-396
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• 2018

The Kori-1 Nuclear Power Plant (NPP), WH 2-Loop Pressurized Water Reactor (PWR) operated for approximately 40 years in Korea, was permanently ceased on June 18, 2017. To reduce worker exposure to radiation by reducing the dose rate in the system before starting main decommissioning activities, the permanently ceased Kori-1 NPP will be subjected to full system decontamination. Generally, the range of system decontamination includes Reactor Pressure Vessels (RPV), Pressurizer (PZR), Steam Generators (SG), Chemical & Volume Control System (CVCS), Residual Heat Removal System (RHRS), and Reactor Coolant System (RCS) piping. In order to decontaminate these systems and equipment in an effective manner, it is necessary to evaluate the influence of the flow characteristics in the RCS during the decontamination period. There are various methods of providing circulating flow rate to the system decontamination. In this paper, the flow characteristics in Kori-1 NPP reactor coolant according to RHR pump operation were evaluated. The evaluation results showed that system decontamination using an RHR pump was not effective at decontamination due first to impurities deposited in piping and equipment, and second to the extreme flow unbalance in the RCS caused deposition of impurities.

• Journal of Energy Engineering
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• v.27 no.4
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• pp.70-79
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• 2018

A heat pump system using the thermal effluent from the Jeju thermal power plant of KOMIPO was constructed with the capacity of 300 RT to supply cool or hot water to greenhouse facilities located 3 km from the power station. The way of transporting heat from the thermal effluent to greenhouses at a long distance was optimized, and a monitoring system to measure the water temperature and detect a leakage in a pipe conduit was also installed. This paper presents the system configuration of the constructed heat pump system for air conditioning and heating of greenhouse facilities in Jeju, and the characteristics of major components deployed in the system. The preoperational tests of the heat pump system were conducted during the summer season in 2018 for evaluation of its cooling performance. The operational stability and cooling performance of the heat pump system were confirmed by investigating the measured fluid temperature and flow rate, and COP of the heat pump in a cooling mode.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.34 no.11
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• pp.1773-1778
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• 2010

To evaluate the structural integrity of the main cooling-water pump of a nuclear power plant under different seismic conditions, the seismic analysis was performed in accordance with IEEE-STD-344 code. The finite element computer program, ANSYS, was used to perform both mode frequency analysis and response spectrum analysis for the pump assembly. The natural frequencies, the mode shapes, and the mode participation factors were obtained from the results of the mode frequency analysis. The stresses resulting from various loadings and their combinations were within the allowable limits specified in the above-mentioned IEEE code. The results of the seismic evaluation fully satisfied the structural acceptance criteria of the IEEE code. Thus, it was proved that the structural integrity of the pump assembly was satisfactory.

• Proceedings of the Korean Nuclear Society Conference
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• 2004.10a
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• pp.299-300
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• 2004