• Nuclear Engineering and Technology
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• 제30권5호
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• pp.403-415
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• 1998

The CANDU 6 reactor assembly consists of an array of 380 pressure tubes, which are installed horizontally in a large cylindrical vessel, the Calandria, containing the low pressure heavy water moderator. The pressure tube is located inside the calandria tube and the annulus between these tubes, which forms a closed loop with $CO_2$ gas recirculating, is called the Annulus Gas System(AGS). It is designed to give an alarm to the operator even for a small pressure tube leak by a very sensitive dew point meter so that he can take a preventive action for the pressure tube rupture incident. To judge whether the operator action time is enough or not in the design of Wolsong 2,3 & 4, the Leak Before Break(LBB) assessment is required for the analysis of the pressure tube failure accident. In order to provide the required data for the LBB assessment of Wolsong Units 2, 3, 4, a series of leak detection capability tests was performed by injecting controlled rates of heavy water vapour. The data of increased dew point and rates of rise were measured to determine the alarm set point for the dew point rate of rise of Wolsong Unit 2. It was found that the response of the dew point depends on the moisture injection rate, $CO_2$ gas flow rate and the leak location. The test showed that CANDU 6 AGS can detect the very small leaks less than few g/hr and dew point rate of rise alarm can be the most reliable alarm signal to warn the operator. Considering the present results, the first response time of dew point to the AGS $CO_2$ flow rate is approximated.

• 대한기계학회논문집A
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• 제31권2호
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• pp.277-284
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• 2007

The steam generator in a nuclear power plant is a large heat exchanger that uses heat from a reactor to generate steam to drive the turbine generator. Rupture of a steam generator tube can result in release of fission products to environment outside. Therefore, an accurate integrity assessment of the steam generator tubes with cracks is of great importance for maintaining the safety of a nuclear power plant. The steam generator tubes are supported at regular intervals by support plates and rotations of the tubes are restrained. Although it has been reported that the limit load for a circumferential crack is significantly affected by boundary condition of the tube, existing limit load solutions do not consider the restraining effect of support plate correctly. In addition, there are no limit load solutions for circumferential cracks in U-bend region with the effect of the support plate. This paper provides detailed limit load solutions for circumferential cracks in top of tube sheet and the U-bend regions of the steam generator tube with the actual boundary conditions to simulate the restraining effect of the support plate. Such solutions are developed based on three dimensional finite element analyses. The resulting limit load solutions are given in a polynomial form, and thus can be simply used in practical integrity assessment of the steam generator tubes.

• Corrosion Science and Technology
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• 제5권4호
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• pp.123-128
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• 2006

It was analyzed the causes of boiler tube rupture due to a degradation and corrosion on the boiler tubes in fossil power plant. The experiments were carried out among samples taken from the operating facilities. The result were analyzed based on experimental results from mechanical strength, microstructure observation, and hardness measurement in order to determine the cause of local rupture on boiler tubes. In general, 2.25Cr-1Mo steel generates carbides, it is coarsened, its ductility and strength abruptly decreased as degradation is in progress, In order to confirm this phenomenon, we observed changes of the mount of Cr and Mo of carbide by carrying out EDX chemical composition analysis. The amount of Cr and Mo in the degraded material or service exposed material gradually increased the amount of Mo but initially they were almost maintained at the same amount. Furthermore, we observed that the carbide become coarsened both in the grain and at the grain boundary. Tensile test was carried out to measure a material hardness and to recognize a drop-off of hardness. Overall result for tensile strength and hardness turned out to be lower than new material and mechanical strength and hardness was degraded as the material degradation was in progress.

• 한국가스학회지
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• 제16권5호
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• pp.28-34
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• 2012

보일러 튜브용 크롬강의 열화기구에 대해 고찰하였다. 크롬강의 열화는 경도 및 응력-파단 수명의 변화를 동반하는데, 이는 주로 탄화물 분해 거동에 의존한다. 그런데 2Cr강인 T22의 열화는 이러한 탄화물 거동에 따르지만 12Cr강인 X20의 열화는 마르텐사이트 래쓰 성장거동에 따르기도 한다. T22와 X20의 열화에 의한 경도값과 응력-파단수명 변화현상을 미세조직적인 측면에서 고찰하였다.

• Journal of Chest Surgery
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• 제9권1호
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• pp.98-101
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• 1976

Clinical observations were made on a total 104 cases of spontaneous pneumothorax during the period of 5 years from MAR. 1971 to MAR. 1976 at the Dept. of Chest Surgery, Korea University Medical College. The ratio of male to female cases was 7.4 : 1 in male predominance, and age distribution was 73% of the total cases within the 15-40 age range. There were 56 patients (53.8%) of right sided pneumothorax and 44 patients (42.3%) of left sided pneumothorax, 8 patients presented bilateral spontaneous pneumothorax. The etiologic factors were tuberculous origin in 42.3%, rupture of bullae or blebs in 15.4%, acute pulmonary infection in 3.8%, pulmonary paragonimiasis in 2.9%, cysticercosis 1%, Marfan's syndrome in 1.9%, unknown causes in 32.7%. In all 104 cases, closed chest tube thoracostomy were performed for reexpansion of collapsed lung. In 50 cases of 104 cases, intrapleural instillation of Quinacrine HCl (Atabrine, Winthrop Laboratories, New York, N.Y.) thru the chest tube was used to control of recurrent pneumothorax, and compared observation with 54 cases control group of closed chest tube thoracostomy only. In both groups, 1) cure rates were 63% in control group and 88% in Atabrine group. 2) average duration of inserted chest tube were 5.6 days in control group and 5.7 days in Atabrine group.

• 대한기계학회논문집A
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• 제20권9호
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• pp.2894-2900
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• 1996

The stream generator tubes represent an integral part of a major barrier against the fission product release to the environment. So, the rupture of these tubes could permit flow of reactor coolant into the secondary system and injure the safety of reactor coolant system. Therefore, if the crack was detected during In-Service Inspection of tubes the cracked tube should be evaluated by the pulgging criteria and plugged or not. In this study, the fracture mechanics evaluation is carried out on the thru-wall axial crack due to Primary Water Stress Corrosion Cracking in the roll transition aone of steam generator tube to help the assurence the integrity of tubes and estabilish the plugging criteria. Due to the Inconel which is used as tube material is more ductile than others, the plastic instability repture theory was used to calculate the critical and allowable crack length. Based on Leak Before Break concept the leak rate for the critical crack length and the allowable leak rate are compared and the safety of tubes was given.

• Journal of Welding and Joining
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• 제28권1호
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• pp.41-46
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• 2010

The low alloy-steel material(1.0Cr-0.5Mo, SA213T12), which has widely been used for the waterwall tube in the conventional power plant, do not have enough creep rupture strength for waterwall tubes of the Ultra-supercritical(USC) boilers. According to this reason, the high-strength low alloy-steel(2.25Cr-1.0Mo, SA213T22) has newly been adopted for the waterwall tube in the USC boilers. This paper presents failure analysis on weld-joint of the waterwall tubes in USC boilers. Visual inspections were performed to find out the characteristics of the fracture. Additionally both microscopic characteristics and hardness test were carried out on failed tube samples. Failures seem to happen mainly because the welding process has not been conducted strictly.(preheating, P.W.H.T and so forth). Thus, this paper has the purpose to describe the main cause of the poor welding process and to explain how to prevent similar failures in those weld-joints.

• 한국추진공학회지
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• 제14권4호
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• pp.10-15
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• 2010

Two-stage light-gas gun은 짧은 시간동안 초고압을 발생시키기 용이하므로, 현재까지 고속충격역학, 발사체 공기역학, 재료역학 등 다양한 공학 분야에서 활용되어 왔다. 일반적으로 2단식 경가스총은 고압실, 압축실 그리고 발사관으로 비교적 간단한 구조로 구성되며, 각 부분은 격막에 의해 구분되어있다. 본 연구는 2단식 경가스총을 초고압 액체 제트 분사에 적용하기 위한 기초적 연구로서, 고압실 하류에 설치된 제1격막의 파막 압력의 변화에 따른 발사체의 속도 변화 및 관내 압력 거동을 조사하였다. 그 결과, 제1격막이 파막 되는 압력은 피스톤의 가속에 큰 영향을 미치며, 피스톤이 가속될수록 고압을 생성하는데 용이하였다.

• Corrosion Science and Technology
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• 제3권1호
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• pp.30-33
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• 2004

It has been reported that some events of a rupture of seam generator tube have occurred in nuclear power plants around the world. Main causes of the leakage are from various types of corrosion in the steam generator(SG) tubings. Primary water stress corrosion cracking(PWSCC) of steam generator tubings have occurred in many tubes in Korean plant, and they were repaired using sleeves or plugs, In order to develop proper repair criteria, it is necessary to ascertain the leak behavior of the tubings. A high pressure leak and burst testing system was manufactured. Various types of Electro Discharged Machined (EDM) notches were developed on the SG tubes. Leak rate and burst pressure were measured on the tubes at room temperature. Burst pressure of the part through wall defected tubes depends on the defect depth, Water flow rates after the burst were independent of the t1aw types; tubes having 20 to 60 mm long EDM notches showed similar flow rates regardless of the defect depth. A fast pressurization rate gave the tube a lower burst pressure than the case of a slow pressurization.

• 한국해양공학회지
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• 제21권2호
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• pp.50-59
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• 2007

Steam generator tubes are degraded from wear, stress corrosion cracking, rupture and fatigue and so on. Therefore, the failure assessment of steam generator tube is very important for the integrity of energy plants. In the steam generator tubes, sometimes, the local wall thinning may result from severe degradations such as erosion-corrosion damage and wear due to vibration. In this paper, the elasto-plastic analysis was performed by FE code ANSYS on steam generator tubes with wall thinning. Also, the four-point bending tests were performed on the wall thinned specimens, and then it was compared with the analysis results. We evaluated the failure mode, fracture strength and fracture behavior from the experiment and FE analysis. Also, it was possible to predict the crack initiation point by estimating true fracture ductility under multi-axial stress conditions at the center of the thinned area from FE analysis.