• 한국압력기기공학회 논문집
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• 제12권1호
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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.

• 한국전기전자재료학회:학술대회논문집
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• 한국전기전자재료학회 2009년도 하계학술대회 논문집
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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

• 한국조명전기설비학회:학술대회논문집
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• 한국조명전기설비학회 2005년도 춘계학술대회논문집
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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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• 제7권2호
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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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• 제56권3호
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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.

• 한국안전학회지
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• 제20권2호
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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.

• 방사성폐기물학회지
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• 제11권1호
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• pp.31-39
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• 2013

핵연료는 원자로 운전 중 예기치 못한 상황에서 연료 결함을 초래할 수 있다. 핵연료 결함은 연료봉의 수소화나 이물질에 의한 금속 마모, 그리고 펠렛과 피복관의 상호작용에 의해 피복관이 손상된다. 이렇게 손상된 핵연료의 결함원인을 규명하는 것은 원자력발전의 안전운전에 중요하다고 사료된다. 핵연료가 손상되면 원자로 냉각재가 오염되어 원자로 출력을 낮추거나, 발전소를 정지할 수도 있다. 모든 사용후연료는 건식저장고로 이동 보관되어야 하나, 결함연료는 이동할 수 없으므로 이 연구의 목적은 중수로형 원자로에서 연료가 인출된 후 사용후연료 저장조에서 보관된 연료에 대하여 결함 여부를 판단할 수 있는 기술을 개발하고자 하였다. 이 연구를 통하여 핵종 누설 검출 기술을 이용한 사용후연료 검사기술을 개발하였으며, 이 기술을 월성발전소에 적용함으로써, 검사기술 및 검사시스템에 대한 성능을 입증하였다.

• 공업화학
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• 제9권2호
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• pp.268-272
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• 1998

액체금속로 냉각재인 액체 소듐에서 시편의 누출특성을 소듐-물 반응 실험에 의해 조사하였다. 소듐-물 반응 현상의 확인은 물 누출 실험 전후에 Fe, Cr 및 Ni 등과 같은 시편의 조성 변화로 확인하였다. $100kg/cm^2$의 누출 압력으로 4시간 동안 시편의 누출 경로를 통해 물을 누출시킨 결과, 누출경로에서 소듐-물 반응생성물들이 침적되어 있는 것을 확인하였으나, 부식에 의해 누출경로가 완전 파열되어 다량의 수증기가 액체 소듐속으로 빠져나가는 re-openning 현상은 관찰되지 않았다. 시편의 누출경로가 막히는 self-plugging 현상은 소듐-물 반응에 의한 반응생성물과 시편의 부식에 의한 부식 생성물이 주 원인으로 추정되고, re-openning 현상은 시편의 누출경로에서 열적인 transient로 추정되었다.

• 대한기계학회논문집B
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• 제34권9호
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• pp.867-873
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• 2010

수냉각 발전기 고정자 권선에서의 냉각수 누수 및 흡습에 의한 절연파괴 손상사례가 국내 및 국외에서 자주 발생되고 있다. 이러한 사고는 막대한 경제적 피해뿐만 아니라 전력의 안정적 공급 측면에서 매우 심각한 계통 사고로 연결될 수 있다. 특히 국내 발전기는 15년 이상 운전되어 열화가 진행된 발전기가 50% 이상이며, 계획예방정비 기간 중에 권선에서의 누설 및 흡습 권선이 종종 발견되고 있다. 기존에는 누수 시험 전 과정인 권선 건조 과정을 무시한 채 누수 시험 결과만으로 권선 누설 여부를 진단하였으나 본 논문에서는 누수 시험을 위한 준비 단계인 진공 건조 시의 권선 내부의 진공도 패턴 분석을 통해 권선 누설 및 흡습 여부를 예측진단할 수 있는 방법을 실험적으로 증명하였다.

• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2003년도 가을 학술논문집
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• pp.90-95
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• 2003

환형기체(Annulus Gas) 계통의 화학적 관리는 계통 재질의 건전성 확보와, 계통내 냉각재 또는 감속재 중수 유입 여부 감지 및 방사선량 저감화 등을 통하여 설비의 안전성과 신뢰성을 유지하는데 목적이 있다. 환형기체 계통의 화학 관리절차서 중 CO_2$ 규격 관리는 계통 재질의 건전성 확보와 방사선량 증가에 직결되기 때문에 고순도의 품질 보증이 요구되고 있다. CO_2$의 순도가 기준 값에 미달될 경우는 계통내에 직접적인 영향을 줄뿐 아니라 주변 환경으로의 오염 가능성도 상존하기 때문에 불순물의 정량관리는 매우 중요하다. 따라서 중수로 환형기체 계통에 공급되는 CO_2$ 중의 C, N_2$ 및 Ar등의 농도분석을 통하여 계통내에서 생성 될 수 있는 방사능 inventory를 평가하였으며 CO_2$의 불순물 관리 최적화와 중수로에서 생성되는 기체 방사성 폐기물 관리에 유용한 정보로 사용 될 수 있을 것이다.