• 한국산학기술학회논문지
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• 제21권10호
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• pp.379-385
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• 2020

본 논문은 전투차량에 장착되는 온수히터의 코어부위 냉각수 누수방지를 위한 설계에 관한 것이다. 온수히터는 가온된 냉각수를 승무원실의 히터 코어에 흘려 난방하는 장치로, 전투차량 운용 시 온수히터 코어 부위의 냉각수 누수현상이 발생하는 문제점이 확인되었다. 이 문제점은 주로 코어의 탱크와 튜브 부위의 접합부에서 발생하였는데, 이 부위가 취약하여 고압을 가압하였을 때 누수가 발생한 것으로 추정하였다. 이 문제를 개선하기 위하여 용접방식을 개선하고 온수히터 코어 말단 부위를 높은 압력에서 견딜 수 있는 구조로 변경하였다. 기존 코어와 개선 코어에 대하여 순차적으로 압력을 가하였을 때 기존 코어는 7.0 kgf/㎠에서 누수가 발생하였으며 개선 코어는 17.0 kgf/㎠까지 견고하게 구조를 유지하여 개선이 되었음을 입증하였다. 마지막으로 개선한 구조의 체계 적합성을 입증하기 위하여 성능시험 및 환경시험을 실시하였다. 본 논문의 연구결과를 바탕으로 제작된 개선 온수히터는 전투차량에 적용될 예정이며, 신뢰성 확보를 통한 방위력 향상과 유사 장비의 설계 및 고장분석에도 참고자료가 될 것으로 기대된다.

• Journal of Radiation Protection and Research
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• 제43권4호
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• pp.154-159
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• 2018

Background: To prevent small leakage accidents, a real-time and direct detection system for small leaks with a detection limit below that of existing systems, e.g. $0.5gpm{\cdot}hr^{-1}$, is required. In this study, a small-size beta detector, which can be installed inside the reactor containment (CT) building and detect small leaks directly, was suggested and its feasibility was evaluated using MCNPX simulation. Materials and Methods: A target nuclide was selected through analysis of radiation from radionuclides in the reactor coolant system (RCS) and the spectrum was obtained via a silicon detector simulated in MCNPX. A window was designed to reduce the background signal caused by other nuclides. The sensitivity of the detector was also estimated, and its shielding designed for installation inside the reactor CT. Results and Discussion: The beta and gamma spectrum of the silicon detector showed a negligible gamma signal but it also contained an undesired peak at 0.22 MeV due to other nuclides, not the $^{16}N$ target nuclide. Window to remove the peak was derived as 0.4 mm for beryllium. The sensitivity of silicon beta detector with a beryllium window of 1.7 mm thickness was derived as $5.172{\times}10^{-6}{\mu}Ci{\cdot}cc^{-1}$. In addition, the specification of the shielding was evaluated through simulations, and the results showed that the integrity of the silicon detector can be maintained with lead shielding of 3 cm (<15 kg). This is a very small amount compared to the specifications of the lead shielding (600 kg) required for installation of $^{16}N$ gamma detector in inside reactor CT, it was determined that beta detector would have a distinct advantage in terms of miniaturization. Conclusion: The feasibility of the beta detector was evaluated for installation inside the reactor CT to detect small leaks below $0.5gpm{\cdot}hr^{-1}$. In future, the design will be optimized on specific data.

• Nuclear Engineering and Technology
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• 제54권7호
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• pp.2748-2754
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• 2022

Because existing reactant coolant system (RCS) leakage detection mechanisms are insensitive to small leaks, a real-time, direct detection system with a detection threshold below 0.5 gpm·hr^-1 was studied. A beta-ray detection system using a silicon detector with good energy resolution for beta rays and a low gamma-ray response was proposed. The detection performance in the leakage condition was evaluated through experiments and simulations. The concentration of ¹⁶N in the coolant corresponding to a coolant leakage of 0.5 gpm was calculated using the analytic method and ORIGEN-ARP. Based on the concentration of ¹⁶N and the measurement of the silicon detector with ⁹⁰Sr/⁹⁰Y, the beta-ray count rate was estimated using MCNPX. To evaluate the effect of gamma rays inside the containment building, the signal-to-noise ratio (SNR) was calculated. To evaluate the count rate ratio, the radiation field inside the containment building was simulated using MCNPX, and response evaluation experiments were performed using beta and gamma rays on the silicon detector. The expected beta-ray count rate at 0.5 gpm leakage was 7.26 × 10⁵ counts/sec, and the signal-to-background count rate ratio exceeded 88 for a transport time of 10 s, demonstrating its suitability for operation inside a reactor containment building.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2004년도 춘계학술대회
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• pp.426-431
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• 2004

Leaking water coolant into stator electrical insulation is a growing concern for the aging water-cooled generator since leaks in the generator water-cooled stator winding can affect machine availability and insulation life. But a domestic techniques of such field are insufficient and depend wholly on GE or TOSHIBA technique. Therefore this paper introduces measuring principle and developed measuring system, which has been used to detecting wet absorption. We accomplished the experiment with a stator promotion of virtue which is used in actual power plant. Also, Experimental method of generator stator winding, which is investigated into wet absorption test.

• 한국전기전자재료학회:학술대회논문집
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• 한국전기전자재료학회 2004년도 하계학술대회 논문집 Vol.5 No.1
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• pp.453-456
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• 2004

This paper surveys the latest findings on vegetable-oil-based dielectric coolants in power systems. In recent years, environmental concerns have been increased on the use of poorly biodegradable mineral oils in distribution and power transformers in regions where spills from leaks and equipment failure could contaminate the surroundings. In addition, there are demands to improve equipment efficiencies in power systems. In this reason, researches were started in the mid 1990s to develop a fully biodegradable dielectric coolants. Vegetable oil was considered the most likely candidate for a fully biodegradable dielectric coolants. Vegetable-oil-based dielectric coolants provide the advantages of high level of biodegradability, renewable natural resource, non-toxic properties, enhanced fire safety, more effective cooling and good dielectric strength for many electrical equipment.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 춘계학술발표회논문집(1)
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• pp.291-295
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• 1996

In CANDU reactor, each pressure tubes contain twelve fuel bundles and provide the inlet and outlet for the primary coolant. If a leak develops in the pressure tube, it is detected by Annulus Gas System which contains circulating dry $CO_2$ gas. Since the leaks caused by the flaws are resulted in pressure tube break, establishment of flaw assessment method is very significant in view of the fracture mechanics. In this paper, various criteria for assessing the flaws are presented to prevent the tube rupture and ensure the integrity of reactor operating.

• 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.

• 한국압력기기공학회 논문집
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• 제12권1호
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• pp.126-133
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• 2016

Recently, the research and development of Sodium-cooled Fast Reactors (SFRs) have made progresses. However, liquid sodium, the coolant of an SFR, is chemically unstable and sodium fire can be occurred when liquid sodium leaks from sodium pipe. To reduce the damage by the sodium fire, many fire walls and fire extinguishers are needed for SFRs. LBB concept in SFR might reduce the scale of sodium fire and decrease or eliminate fire walls and fire extinguishers. Therefore, LBB concept can contribute to improve economic efficiency and to strengthen defense-in depth safety. The LBB assessment procedure has been well established, and has been used significantly in light water reactors (LWRs). However, an LBB assessment of an SFR is more complicated because SFRs are operated in elevated temperature regions. In such a region, because creep damage may occur in a material, thereby growing defects, an LBB assessment of an SFR should consider elevated temperature effects. The procedure and method for this purpose are provided in RCC-MRx A16, which is a French code. In this study, LBB assessment was performed for PGSFR IHTS hot-leg pipe according to RCC-MRx A16 and the applicability of the code was discussed.

• 대한기계학회논문집A
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• 제37권3호
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• pp.387-395
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• 2013

최근 국외의 원자로 상부헤드 CRDM 관통노즐에 일차수 응력부식균열로 인한 냉각수 누출사고가 발생하였다. 일차수응력부식균열은 부식에 민감한 재료, 인장 잔류 응력 및 부식 환경 등의 3 가지 요인의 상호작용에 의해 발생하는 것으로 알려져 있기 때문에 응력 부식 균열 발생 및 균열 진전을 억제하기 위해서는 용접에 의한 잔류응력에 대한 정확한 예측이 선행되어야 한다. 본 논문에서는 국내 Westinghouse 형 원자로 상부 헤드 관통노즐(CRDM)을 대상으로 노즐의 두께 및 형상 비($r_o/t$)에 따른 노즐 잔류응력 분포 특성에 대해 연구를 수행하였다. 국내에 현존하는 원자로 상부헤드 관통노즐의 실제크기($r_o$=51.6, t=16.9 mm)를 기준으로 노즐의 두께 및 형상 비($r_o/t$=2, 3, 4)의 변수를 정립하였으며 정중앙 및 최외곽에 위치한 노즐을 대상으로 연구를 수행하였다.

• 방사선산업학회지
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• 제9권4호
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• pp.217-221
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• 2015

When reactor coolant leaks occur due to cracks of a steam generator's tube, radioactive materials contained in the primary cooling water in nuclear power plant are forced out toward the secondary systems. At this time the secondary water purification resin in the ion exchange resin tower of the steam generator blowdown system is contaminated by the radioactivity of the leaked radioactive materials, so we pack this in special containers and store temporarily because we could not dispose it by ourselves. If steam generator tube leakage occurs, it produces contaminated spent resins annually about 5,000~7,000 liters. This may increase the amount of nuclear waste productions, a disposal working cost and a unit price of generating electricity in the plant. For this reasons, it is required to develop a decontamination process technique for reducing the radioactive level of these resins enough to handle by the self-disposal method. In this research, First, Investigated the structure and properties of the ion exchange resin used in a steam generator blowdown system. Second, Checked for a occurrence status of contaminated spent resin and a disposal technology. Third, identified the chemical characteristics of the waste radionuclides of the spent resin, and examined ionic bonding and separation mechanism of radioactive nuclear species and a spent resin. Finally, we carried out the decontamination experiment using chemicals, ultrasound, microbubbles, supercritical carbon dioxide to process these spent resin. In the case of the spent resin decontamination method using chemicals, the higher the concentration of the drug decontamination efficiency was higher. In the ultrasound method, foreign matter of the spent resin was removed and was found that the level of radioactivity is below of the MDA. In the microbubbles method, we found that the concentration of the radioactivity decreased after the experiment, so it can be used to the decontamination process of the spent resin. In supercritical carbon dioxide method, we found that it also had a high decontamination efficiency. According to the results of these experiments, almost all decontamination method had a high efficiency, but considering the amounts of the secondary waste productions and work environment of the nuclear power plant, we judged the ultrasound and supercritical carbon dioxide method are suitable for application to the plant and we established the plant applicable decontamination process system on the basis of these two methods.