• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.12 no.1
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• pp.93-100
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• 2016

Since the sodium-cooled fast reactor is operated in a hostile environment due to the use of liquid sodium as its coolant, advanced techniques for in-service inspection are required to periodically verify the integrity of the reactor. This paper presents the development of in-service inspection techniques for Proto-type Generation IV Sodium-cooled Fast Reactor. First, the 10 m long plate-type ultrasonic waveguide sensor has been developed for in-service inspection of reactor internals, and its feasibility was verified through several under-water and under-sodium experiments. Second, the combined inspection system for in-service inspection of ferromagnetic steam generator tubes has been developed. The remote field eddy current testing and magnetic flux leakage testing can be conducted simultaneously by using the developed inspection system, and the detectability was demonstrated through several damage detection experiments. Finally, the electro-magnetic acoustic transducer which can withstand high temperature and be installable in the remote operated vehicle has been developed for in-service inspection of the reactor vessel, and its detectability was investigated through damage detection experiments.

• Proceedings of the Korean Institute of Electrical and Electronic Material Engineers Conference
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• 2009.06a
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• pp.361-361
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• 2009

KEPCO High Temperature Superconducting (HTS) cable system rated with $3\Phi$, 22.9kV, 1250A was laid in 2006, and the long term test is in progress. The HTS cable system with the cooling system has been operated below cryogenic temperature. That environment exposes the system to the thermo-mechanical stress due to the significant temperature difference, and the cooling system has moving parts for the forced circulation of the coolant. Therefore the HTS cable system experiences thermal fatigue and moving part such as liquid nitrogen pump need a regular replacement every 5000 hours Building the assessment criterion, the maintenance procedure was established and regular preventive maintenance was done; improvement of the termination structure and the replacement of the bearing of liquid nitrogen pump. Following the proper process, the reliability assessment test including He leakage detection and the stability of flow rate was performed. This paper describes the process and result of the first regular maintenance of KEPCO HTS cable system

• Proceedings of the Korean Institute of IIIuminating and Electrical Installation Engineers Conference
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• 2005.05a
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• pp.385-390
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• 2005

This paper provided a counter measures against the troubles and accidents that are likely to take place in the power plant using hydrogen gas as a coolant for the cooling system of the generator. Because of the extremely wide flammability limits of hydrogen in comparison to the other flammable gases, the safety measures against the hydrogen accidents is very important to ensure the normal operation of electric-power facility. This study's purpose was a presentation of standard model of safety management of hydrogen equipments in the coal firing power plant such as following items: 1) providing the technical prevention manual of the hydrogen explosions and hydrogen fires occurring in the cooling system of power generator; 2) the selection of explosion-proof equipments in terms of the risk level of operating environment; 3) the establishment of regulations and counter measures, such as the incorporation of gas leakage alarm device, for preventing the accidents from arising; 4) the establishment of safety management system to ensure the normal operation of the power plant.

• Journal of Electrical Engineering and Technology
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• v.7 no.2
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• pp.230-235
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• 2012

Cases of insulation breakdown damage of water-cooled generator stator windings occur frequently due to coolant leakage and water absorption worldwide. Such serious accidents may cause not only enormous economic loss but also very serious grid accidents in terms of stable supply of electric power. More than 50 % of domestic generators have been operated for more than 15 years, and leak and water absorption problem of windings are often found during the planned preventive maintenance period. Since 2005, leak and water absorption tests have been performed for total watercooled stator windings after fully drying the inside of the windings. The results are then comprehensively analyzed. The result of the test performed by GE, a foreign manufacturer, for 141 generators showed failures in 80 of them (failure rate: 57 %), whereas in the tests carried out in Korean domestic power plants, only 14 out of 50 generators showed failures (failure rate: 28 %).

• Nuclear Engineering and Technology
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• v.56 no.3
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• pp.832-845
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• 2024

Reactor cavity cooling systems (RCCSs) comprising passive safety features use the atmosphere as a coolant, which cannot be lost. However, their drawback is that they are easily affected by atmospheric disturbances. To realize the commercial application of the two types of passive RCCSs, namely RCCSs based on atmospheric radiation and atmospheric natural circulation, their safety must be evaluated, that is, they must be able to remove heat from the reactor pressure vessel (RPV) surface at all times and under any condition other than under normal operating conditions. These include both expected and unexpected natural phenomena and accidents. Moreover, they must be able to eliminate the heat leakage emitted from the RPV surface during normal operation. However, utilizing all of the heat emitted from the RPV surface increases the degree of waste heat utilization. This study aims to understand the characteristics and degree of passive safety features for heat removal by comparing RCCSs based on atmospheric radiation and atmospheric natural circulation under the same conditions. It was concluded that the proposed RCCS based on atmospheric radiation has an advantage in that the temperature of the RPV could be stably maintained against disturbances in the ambient air.

• Journal of the Korean Society of Safety
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• v.20 no.2 s.70
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• pp.32-37
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• 2005

The purpose of this study is for the sensitivity study f9r a Steam Generator Tube Rupture (SGTR) of the System-integrated Modular Advanced ReacTor for a Pilot (SMART-P) plant. The thermal hydraulic analysis of a SGIR for the Limiting Conditions for Operation (LCO) is performed using TASS/SMR code. The TASS/SMR code can calculate the core power, pressure, flow, temperature and other values of the primary and secondary system for the various initiating conditions. The major concern of this sensitivity study is not the minimum Critical Heat Flux Ratio(CHFR) but the maximum leakage amount from the primary to secondary sides at the steam generator. Therefore the break area causing the maximum accumulated break flow is researched for this reason. In the case of a SGIR for the SMART-p, the total integrated break flow is 11,740kg in the worst case scenario, the minimum CHFR is maintained at Over 1.3 and the hottest fuel rod temperature is below 606"I during the transient. It means that the integrity of the fuel rod is guaranteed. The reactor coolant system and the secondary system pressures are maintained below 18.7MPa, which is system design pressure.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.11 no.1
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• pp.31-39
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• 2013

Nuclear fuel can be damaged under unexpected circumstances in a nuclear reactor. Fuel rod failure can be occurred due to debris fretting or excessive hydriding or PCI (Pellet-to-clad Interaction) etc. It is important to identify the causes of such failed fuel rods for the safe operation of nuclear power plants. If a fuel rod failure occurs during the operation of a nuclear power plant, the coolant water is contaminated by leaked fission products, and in some case the power level of the plant may be lowered or the operation stopped. In addition, all spent fuels must be transferred to a dry storage. But failed fuel can not be transferred to a dry storage. Therefore, the purpose of this study is to develop a system which is capable of inspecting whether the spent fuel in the storage pool is failed or not. The sipping technology is to analyze the leakage of fission products in state of gas and liquid. The failed fuel inspection system with gamma analyzer has successfully demonstrated that the system is enough to find the failed fuel at Wolsong plant.

• Applied Chemistry for Engineering
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• v.9 no.2
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• pp.268-272
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• 1998

Water leak phenomena in the liquid sodium which is a coolant of liquid metal reactor, were investigated by carrying out sodium-water reaction experiment. It was confirmed that sodium and water react each other by the analysis of material composition of aspecimen at the end of experiment. When steam of $100kg/cm^2$ was passed through the leak path of the specimen for 4 hours, reaction products from sodium-water reaction were observed on the leak site. However, re-opening phenomena were not observed at this condition. It was interpretted that the reaction product precipitated on leak path and thermal transient caused self-plugging and re-openning phenomena, respectively.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.34 no.9
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• pp.867-873
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• 2010

The failure of water-cooled generator stator windings as a result of insulation breakdown due to coolant water leaks and water absorption often occurs worldwide. Such failure can cause severe grid-related accidents as well as huge economic losses. More than 50% of domestic generators have been operated for over 15 years, and therefore, they exhibit signs of aging. Leaking and water-absorbing windings are often found during an overhaul. In an existing method for evaluating the integrity of generator stator windings, the drying process of the interior of the windings is ignored and only final leak tests are performed. In this study, it is shown that water leaks and water absorption in stator windings can be detected indirectly through vacuum pattern analysis in the vacuum drying mode, which is the used in the preparation stage of the leak test.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2003.11a
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• pp.90-95
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• 2003

Chemical management of annulus gas system is carried out for the purpose of ensuring the safety and reliability of the system via securing the integrity of the system, detecting the D$_2$O in-leakage of coolant and/or moderator, and reducing the radiation dose. Since the quality of CO_2$ gas, which is used as a filling gas for annulus gas system at CANDU plants, has a propound effect on the integrity of the system material and the radiation dose, CO_2$ gas of high quality is needed. If the quality of CO_2$ gas does not meet the specification, it may give rise to undesirable effect not only on the annulus gas system, but also on the environment due to the production of radioactive nuclei. Therefore, it is very important to check the impurities of CO_2$ gas. Based on this background, the inventories of C-14 and Ar-41 in CO_2$ gas that is supplied as annulus gas were estimated using the data on concentrations of the impurities of $CO_2$ such as C, N_2$ and Ar. The results of this study is expect to give useful information on optimization of CO_2$ impurities maintenance and management of gaseous radioactive wastes produced at CANDU plants.