• Nuclear Engineering and Technology
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• v.53 no.4
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• pp.1127-1133
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• 2021

A Spent Fuel Pool (SFP) is designed to store spent fuel assemblies in the pool. And, a SFP cooling and cleanup system cools the SFP coolant through a heat exchanger which exchanges heat with component cooling water. If the cooling system fails or interfacing pipe (e.g., suction or discharge pipe) breaks, the cooling function may be lost, probably leading to fuel damage. In order to prevent such an incident, it is required to properly cool the spent fuel assemblies in the SFP by either recovering the cooling system or injecting water into the SFP. Probabilistic safety assessment (PSA) is a good tool to assess the SFP risk when an initiating event for the SFP occurs. Since PSA has been focused on reactor-side so far, it is required to study on the framework of PSA approach for SFP and identify the key factors in terms of fuel damage frequency (FDF) through a case study. In this study, therefore, a case study of SFP-PSA on the basis of design information of APR-1400 has been conducted quantitatively, and several sensitivity analyses have been conducted to understand the impact of the key factors on FDF.

• The KSFM Journal of Fluid Machinery
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• v.12 no.5
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• pp.7-12
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• 2009

A nuclear fuel test loop(after below, FTL) is installed in the IRI of an irradiation hole in HANARO for testing the neutron irradiation characteristics and thermo hydraulic characteristics of a fuel loaded in a light water power reactor or a heavy water power reactor. There is an in-pile section(IPS) and an out-pile section(OPS) in this test loop. When HANARO is operated normally, the fuel loaded into the IPS has a nuclear reaction heat generated by a neutron irradiation. To remove the generated heat and to maintain the operation conditions of the test fuel, a main cooling water system(MCWS) is installed in the OPS of the FTL. The MCWS is composed of a main cooler, a pressurizer, two circulation pumps, a main heater, an interconnection pipe line and instruments. The interconnection pipeline is a closed loop which is connected to an inlet and an outlet of the IPS respectively. The MCWS is under a cold function test during a start-up period. This paper describes the system flow network analysis results of the flow control of a main cooling water system in the HANARO fuel test loop. It was confirmed through the results that the flow was met the system design requirements.

• Transactions of the Korean Society of Mechanical Engineers
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• v.16 no.11
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• pp.2126-2135
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• 1992

In order to prevent the decay heat removal system from failure due to air entrainment or free surface vortex in the piping system, a set of simulating experiments for the midloop operation of nuclear power plant was performed. Through these experiments, a relation between the dimensionless numbers, such as submergence H/d, froude number, reynolds number, was found. However, the effect of reynolds number was negligible for the operation conditions of Nuclear power plant. It was also found that the perturbation of the system by the disturbance such as pump start, valve operation, etc., has a strong effect on the free surface vortex. Furthermore, from a view point of reactor safery, a modified inlet device of reducer type is strongly recommendable for the prevention of air entrainment.

• Proceedings of the KSME Conference
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• 2004.04a
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• pp.72-77
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• 2004

In a primary reactor cooling system(RCS), a dissimilar weld zone exists between cast stainless steel(CF8M) in a pipe and low-alloy steel(SA508 cl.3) in a nozzle. Thermal aging is observed in CF8M as the RCS is exposed for a long period of time to a reactor operating temperature between 290 and $330^{\circ}C$, while no effect is observed in SA508 cl.3. The specimens are prepared by an artificially accelerated aging technique maintained for 300, 1800 and 3600 hrs at $430^{\circ}C$, respectively. The specimens for elastic-plastic fracture toughness tests are prepared one type, which notch is created in the heat affected zone(HAZ) of CF8M. And, the specimens for fatigue crack growth tests are prepared in three classes, which notches are created at the center of deposited zone, the HAZ of CF8M, and the HAZ of SA508 cl.3. From the experiments, the J-integral values with the increase of aging time decrease, and the differences of the fatigue crack growth behaviors are relatively small in the three classes specimens.

• Proceedings of the KSME Conference
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• 2001.06a
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• pp.636-643
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• 2001

Interference between the crossover leg of the reactor coolant system (RCS) and the pipe whip restraints (PWR) has brought a degradation issue of the integrity of the Reactor Coolant System in Westinghouse type nuclear power plants (NPPs) of Korea. According to the gap Inspect ion carried out during planned overhaul (Year 2000), interference between the crossover leg and the PWR was found in each RCS loop. This plant has had the high vibration problem on the RC pump 'B'. The reason for the high vibration in the RC pump 'B' had been massively surveyed and it was found that the crossover leg of RCS contacted with the PWR in hot condition. Since the contact between the crossover leg and the PWR changes the dynamic characteristics of the piping system for the RCS, this is considered as one reason for the high vibration. And a possibility of overstress on the crossover leg due to the contact with the PWR should be evaluated. Through performing RCS integrity analyses, subsequent actions were initiated to increase the gap between those parts. As the results of the appropriate separation between two parts, it was reported that there was no unusual noise or vibration during plant heat-up. In this paper, the evaluations for the gap between the crossover leg and the PWR and the structural integrity due to loop binding is described.

• Nuclear Engineering and Technology
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• v.55 no.12
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• pp.4685-4694
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• 2023

The ability to calculate the material density sensitivity coefficients of power with respect to the material density has broad application prospects for accelerating Monte Carlo-Thermal Hydraulics iterations. The second-order material density sensitivity coefficients for the general Monte Carlo score have been derived based on the differential operator sampling method in this paper, and the calculation of the sensitivity coefficients of cell power scores with respect to the material density has been realized in continuous-energy Monte Carlo code RMC. Based on the power-density sensitivity coefficients, the sensitivity coefficients of power scores to some other physical quantities, such as power-boron concentration coefficients and power-temperature coefficients considering only the thermal expansion, were subsequently calculated. The effectiveness of the proposed method is demonstrated in the power-density coefficients problems of the pressurized water reactor (PWR) moderator and the heat pipe reactor (HPR) reflectors. The calculations were carried out using RMC and the ENDF/B-VII.1 neutron nuclear data. It is shown that the calculated sensitivity coefficients can be used to predict the power scores accurately over a wide range of boron concentration of the PWR moderator and a wide range of temperature of HPR reflectors.

• Nuclear Engineering and Technology
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• v.56 no.6
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• pp.2332-2342
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• 2024

Lead-cooled fast reactors (LFRs) have a wide range of application scenarios, which require the thermal-hydraulic characteristics of LFRs to be reliable. In the present paper, the Lead-cooled fast reactor Thermal-Hydraulic Analysis Code LETHAC was developed, including the models of pipe, heat exchanger, and pool. To verify the correctness of LETHAC, two experimental facilities and three experimental cases were selected, including GFT and PLOFA tests for NACIE-UP and Test-1 for CIRCE. The calculated results show the same and consistent trend with the experimental data, but there are some discrepancies. It can be found that LETHAC is suitable and reliable in predicting the transient behavior of lead-cooled system.

• Nuclear Engineering and Technology
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• v.50 no.7
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• pp.1079-1087
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• 2018

There are many parameters that affect the natural circulation flow, such as height difference, heating power size, pipe diameter, system pressure and inlet temperature and so on. In general analysis the heating power is often regarded as a uniform distribution. The ANSYS-CFX numerical analysis software was used to analyze the flow heat transfer of supercritical water under different heating power distribution conditions. The distribution types of uniform, power increasing, power decreasing and sine function are investigated. Through the analysis, it can be concluded that different power distribution has a great influence on the flow of natural circulation if the total power of heating is constant. It was found that the peak flow of supercritical water natural circulation is maximal when the distribution of heating power is monotonically decreasing, minimal when it is monotonically increasing, and moderate at uniform or the sine type of heating. The simulation results further reveal the supercritical water under different heat transfer conditions on its flow characteristics. It can provide certain theory reference and system design for passive residual heat removal system about supercritical water.

• Proceedings of the Korea Concrete Institute Conference
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• 1999.10a
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• pp.329-332
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• 1999

The crack of concrete induced by the heat of hydration is a serious problem, particularly in concrete structures such as underground box structure, mat-slab of nuclear reactor buildings, dams or large footings, foundations of high rise buildings, etc.. As a result of the temperature rise and restriction condition of foundation, the thermal stress which may induce the cracks can occur. Therefore the various techniques of the thermal stress control in massive concrete have been widely used. One of them is prediction of the thermal stress, besides low-heat cement which mitigates the temperature rise, pre-cooling which lowers the initial temperature of fresh concrete with ice flake, pipe cooling which cools the temperature of concrete with flowing water, design change which considers steel bar reinforcement, operation control and so on. The objective of this paper is largely two folded. Firstly we introduce the cracks control technique by employing low-heat cement mix and thermal stress analysis. Secondly it show the application condition of the cracks control technique like sidewall of LNG in Inchonl.

• Proceedings of the Korea Concrete Institute Conference
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• 1999.04a
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• pp.293-298
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• 1999

The crack of concrete induced by the heat of hydration is a serious problem, particularly in concrete structures such as underground box structure, mat-slab of nuclear reactor buildings, dams or large footings, foundations of high rise buildings, etc.. As a result of the temperature rise and restriction condition of foundation, the thermal stress which may induce the cracks can occur. Therefore the various techniques of the thermal stress control in massive concrete have been widely used. One of them is prediction of the thermal stress, besides low-heat cement which mitigates the temperature rise, pre-cooling which lowers the initial temperature of fresh concrete with ice flake, pipe cooling which cools the temperature of concrete with flowing water, design change which considers steel bar reinforcement, operation control and so on. The objective of this paper is largely two folded. Firstly we introduce the cracks control technique by employing low-heat cement mix and thermal stress analysis. Secondly it show the application condition of the cracks control technique like the subway structure in Seoul.