• Nuclear Engineering and Technology
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• v.44 no.7
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• pp.735-744
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• 2012

In order to quantify the flow distribution characteristics of APR+ reactor, a test was performed on a test facility, ACOP ($\underline{A}$PR+ $\underline{C}$ore Flow & $\underline{P}$ressure Test Facility), having a length scale of 1/5 referring to the prototype plant. The major parameters are core inlet flow and outlet pressure distribution and sectional pressure drops along the major flow path inside reactor vessel. To preserve the flow characteristics of prototype plant, the test facility was designed based on a preservation of major flow path geometry. An Euler number is considered as primary dimensionless parameter, which is conserved with a 1/40.9 of Reynolds number scaling ratio. ACOP simplifies each fuel assembly into a hydraulic simulator having the same axial flow resistance and lateral cross flow characteristics. In order to supply boundary condition to estimate thermal margins of the reactor, the distribution of inlet core flow and core exit pressure were measured in each of 257 fuel assembly simulators. In total, 584 points of static pressure and differential pressures were measured with a limited number of differential pressure transmitters by developing a sequential operation system of valves. In the current study, reactor flow characteristics under the balanced four-cold leg flow conditions at each of the cold legs were quantified, which is a part of the test matrix composing the APR+ flow distribution test program. The final identification of the reactor flow distribution was obtained by ensemble averaging 15 independent test data. The details of the design of the test facility, experiment, and data analysis are included in the current paper.

• Computational Structural Engineering
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• v.10 no.4
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• pp.155-164
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• 1997

A reactor pressure vessel, which contains fuel assemblies and reactor vessel internals, has the thermal stress resulting from the cool-down and heat-up of the vessel wall in combination with the pressure stress from system pressure resulting in large stresses. The combination of the pressure stress and thermal stress along with a decrease in fracture toughness may cause through-wall propagation of a relatively small crack. Therefore, it is necessary to define the relations between operating pressure and temperature during cool-down and heat-up. In this study, theory of fracture mechanics for a pressure/temperature limit curve is investigated and a numerical procedure for generating it is developed. Plant-specific limit curves for the Kori unit 1 plant, the oldest nuclear power plant in Korea, have been obtained for several cooling and heating rates and their results are discussed.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.9 no.1
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• pp.20-24
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• 2013

Reactor Vessel Internals(RVIs), which are installed within the reactor pressure vessel and support the fuel assembly, take responsibility for safety of reactor core. In operating Nuclear Power Plants(NPPs), the RVIs have been exposed to severe conditions such as neutron irradiation, high temperature, high pressure, and high velocity of coolant flow and have degraded by materials aging with long-term operation. Therefore, the effective aging management plan and the appropriate regulatory requirements are necessary to maintain the integrity of RVIs. The purpose of this paper is to provide a review guide for Periodic Safety Review(PSR) of RVIs in presurized water reactor. The review guide is developed based on the revised review guides and reports established from IAEA and USNRC, and the analysis results of design characteristics, aging mechanisms, and operating experiences of RVIs in domestic and international NPPs. Consequently, the developed review guide for PSR of RVIs is expected to contribute an overall strategy and standard for the PSR of RVIs.

• Nuclear Engineering and Technology
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• v.50 no.5
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• pp.654-664
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• 2018

Self-pressurization analysis of the natural circulation integral nuclear reactor through a new dynamic model is studied. Unlike conventional pressurized water reactors, this reactor type controls the system pressure using saturated coolant water in the steam dome at the top of the pressure vessel. Self-pressurization model is developed based on conservation of mass, volume, and energy by predicting the condensation that occurs in the steam dome and the flashing inside the chimney using the partial differential equation. A simple but functional model is adopted for the steam generator. The obtained results indicate that the variable measurement is consistent with design data and that this new model is able to predict the dynamics of the reactor in different situations. It is revealed that flashing and condensation power are in direct relation with the stability of the system pressure, without which pressure convergence cannot be established.

• Nuclear Engineering and Technology
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• v.34 no.4
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• pp.344-357
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• 2002

The present paper reports spontaneous steam explosions observed in fuel coolant interaction experiments using prototypic reactor materials. Pure ZrO$_2$ and a mixture of UO$_2$ and ZrO$_2$ are used. A high temperature molten material in the form of a jet is poured into a subcooled water pool located in a pressure vessel. An induction skull melting technique is used for the melting of the reactor material. In both tests using pure ZrO$_2$ and a mixture of UO$_2$ and ZrO$_2$, either a quenching or a spontaneous steam explosion was observed. The morphology of debris and pressure profile clearly indicate the differences between the qunching cases and explosion cases. The dynamic pressure. dynamic impulse, water temperature, melt temperature, and static pressure Inside the containment chamber were measured . As the spontaneous steam explosion for the reactor material is firstly observed in the present experiments, the results of present experiments could be a siginificant step forward the understanding the explosion of the reactor material.

• Journal of the Korean Society of Combustion
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• v.21 no.4
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• pp.39-47
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• 2016

The aims of this research were to investigate combustion characteristics of lab-scale pressurized oxy-fuel combustion(POFC) system. In this study, the reactor, 800 mm long, was equipped with co-axial burner. Low calorific value syngas that is composed of mainly CO and $H_2$ was used as fuel whereas pure oxygen was used as an oxidant. Thermal heat input to the reactor varied from 2.6 kW to 6.1 kW. The reactor pressure also increases from atmospheric up to 15 bar. The results show that as the pressure increase, the temperature of reactor decreases on the whole in all cases. A significant temperature drop was observed especially at the bottom section of the reactor that exist flame. In addition, the flame instability increases as the pressure increases. Furthermore $NO_x$ emissions increases from atmospheric up to 2 bar. However beyond 2 bar, $NO_x$ emission reduces as pressure increases. Lastly $NO_2$ ratio in $NO_x$ also increases as pressure increases.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.16 no.1
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• pp.84-91
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• 2020

The APR1400 reactor stud holes can be stuck due to high temperatures, high pressure, prolonged engagement, and load changes according to pressure changes in the reactor. Threaded surfaces of a stud hole should be cleaned for the sealing of pressure in reactor vessel by removing any foreign materials which may exist in the stud holes. Human workers can access to the stud hole for the cleaning of stud holes manually, but the radiation exposure of human workers is increased. Robot is an effective way to work in hazardous area. So we introduced robot for the cleaning of stud holes. Localization of mobile robots is generally based on odometry, but with increased mileage, position errors can be accumulated. In order to eliminate cumulative error and to ensure stability of its driving, laser sensors and new control algorithm were utilized. The distance between the robot and the wall was measured by laser sensors, and the control algorithm was implemented so as to travel the desired trajectory by using the measured values from sensors. The performance of driving and hole sensing were verified through field application, and mobile robot was confirmed to be applicable to the APR 1400 NPP.

• Nuclear Engineering and Technology
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• v.55 no.5
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• pp.1604-1615
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• 2023

In an open-pool type research reactor with a downward forced flow in the core, pipes can be under sub-atmospheric pressure because of the large pressure drop at the reactor core in the atmospheric pool. Sub-atmospheric pressure can result in air inflow into the pipe from the pressure difference between the atmosphere and the inside of the pipe, which in a postulated pipe break scenario can lead to the breakdown of the cooling pump. In this study, a plant-scale experiment was conducted to study air inflow in large piping systems by considering the actual operational conditions of an advanced research reactor. The air inflow rate was measured, and the entrained air was visualized to investigate the behavior of air inflow and flow regime depending on the pipe break size. In addition, the developed drift-flux model for a large vertical pipe with a diameter of 600 mm was compared with other correlations. The flow regime transition in a large vertical pipe under downward flow was also studied using the newly developed drift-flux model. Consequently, the characteristics of two-phase flow in a large vertical pipe were found to differ from those in small vertical pipes where liquid recirculation was not dominant.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.7 no.3
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• pp.15-22
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• 2011

The chemistry factor and RTPTS margin for domestic reactor pressure vessel materials were analyzed by using the surveillance data which have been obtained from 8 nuclear power plants in Korea. The surveillance data have been used to assess the integrity of the pressure vessel under the pressurized thermal shock (PTS) event. The chemistry factor, which is determined by the Cu and Ni contents of vessel materials, is considered a proper tool to assess the $RT_{PTS}$. The chemistry factors, which were obtained from the surveillance data of domestic reactor pressure vessels, were investigated and compared with those of Regulatory Guide 1.99 in this study. Regressions for ${\Delta}RT_{NDT}$ were performed to expect the chemistry factor as a function of Cu and Ni, and to estimate $RT_{PTS}$ margin. The margin analysis was performed by comparing the regression graphs and standard deviations with those of Regulatory Guide 1.99. The standard deviations calculated by using the domestic surveillance data for base metal and welds are almost same as the standard deviations which are suggested on Regulatory Guide 1.99, Rev.2.

• Proceedings of the KSME Conference
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• 2008.11a
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• pp.727-728
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• 2008

The objective of this study is to evaluate the integrity for a reactor pressure vessel under the pressurized thermal shock by applying the probability fracture mechanics. A semi-elliptical axial crack is assumed to be in the beltline region of the reactor pressure vessel. The selected random variables are the neutron fluence on the vessel inside surface, the content of copper, nickel, and phosphorus in the reactor pressure vessel material, and initial RTNDT. The probabilistic integrity analysis was performed using the Monte Carlo simulation.