• Proceedings of the Korean Nuclear Society Conference
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• 1996.05b
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• pp.48-54
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• 1996

가압경수로의 부분충수 운전중 RHR 계통의 기능상실시 사고완화를 위해 가압기 manway와 증기 발생기 출구공동 manway를 동시에 개방한 경우에 대한 실험결과를 CATHAHR2 코드를 이용하여 해석하였다. 해석을 통해 이 경우에 발생하는 물리적 현상을 이해하고 이와 같은 과도기에 대해 코드 예측능력을 평가하므로 써, 실제 원전에서 사고시 적절한 사고대응 방안을 강구하는데 참고가 될 수 있도록 해석적 근거를 제시하고자 한다. 연구결과 CATHARE2 코드는 실험을 통해 관측된 주요 물리적 현상들을 타당하게 예측하였으나, 가압기와 밀림관의 DP를 과대 예측하여 원자로 상부공동의 최대압력을 실험보다 약 7kPa 높게 예측하였다. 노심 노출시간도 노심에서 기포율 분포를 비현실적으로 예측하여 실험보다 약 500초 지연되었다. 실험과 코드의 모의결과를 통하여 노심 노출은 중력주입에 의한 냉각수 보충만으로 충분히 회복될 수 있음을 확인하였다. CATHARE2 코드는 비록 상세한 현상들에 대해 다소 불확실성을 내포하였으나, 전반적인 거동분석에는 타당한 것으로 판단된다. CATHARE 코드는 노심에서 계면 마찰력을 줄임으로써 노심의 차압을 개선할 수 있었고, guide 튜브의 위치를 고온관 중심선과 일치시켜 guide 튜브내 액체의 hold-up 기간을 개선할 수 있었으며, 가압기의 계면 마찰력을 증가시켜서 밀림관에서 "plug and clearing" 현상을 모의할 수 있었다.모의할 수 있었다.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.11 no.2
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• pp.68-74
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• 2015

Time domain response analysis for vibro-impact nonlinear behavior of multi-span tube with loose supports was performed using commercial FEA code and user subroutine. Support geometry of multi-span tube with a finite gap is realistically modeled by analytical rigid surface. Model of hydrodynamic force is based on the Qusai-steady model which accounts for the inclined angle of relative flow velocity and time delay between flow force and resulting tube motion. During tube vibration from flow loading, impact and friction at the support location is simulated using commercial FEA code with master slave contact algorithm. Analysis results has reasonable agreement with those of references and test experience. Plan of further refinement of analysis model and future test verification is briefly introduced.

• Transactions of the Korean Society for Noise and Vibration Engineering
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• v.12 no.4
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• pp.261-271
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• 2002

The fluid-elastic instability analysis of the U-tube bundle inside the steam generator is very important not only for detailed design stage of the SG but also for the change of operating condition of the nuclear powerplant. However the calculation procedure for the fluid-elastic instability was so complicated that the consolidated computer program has not been developed until now. In this study, the numerical calculation procedure and the computer program to obtain the stability ratio were developed. The thermal-hydraulic data in the region of secondary side of steam generator was obtained from executing the ATHOS3 code. The distribution of the fluid density can be calculated by using the void fraction, enthalpy, and operating pressure. The effective mass distribution along the U-tube was required to calculate natural frequency and dynamic mode shape using the ANSYS ver. 5.6 code. Finally, stability ratios for selected tubes of the CE type steam generator were computed. We considered the YGN 3.4 nuclear powerplant as the model plant, and stability ratios were investigated at the flow exit region of the U-tube. From our results, stability ratios at the central and the outside region of the tube bundle are much higher than those of other region.

• Proceedings of the KSME Conference
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• 2000.04a
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• pp.25-30
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• 2000

Residual stresses induced in U-bending and tube-to-tubesheet joining processes of PWR's steam generator row-1 tube were measured by X-ray method and Hole-Drilling Method(HDM). The stresses resulting from the Internal pressure and the temperature gradient in the steam generator were also estimated theoretically. In U-bent lesions, the residual stresses at extrados were induced with compressive stress(-), and its maximum value reached -319 MPa in axial direction at ${\psi}=0^{\circ}$ in position. Maximum tensile residual stress of 170MPa was found to be at the flank side at Position of${\psi}=90^{\circ}$, i.e., at apex region. In tube-to-tubesheet fouling methods, the residual stresses induced by the explosive joint method were found to be lower than that by the mechanical roll method. The gradient of residual stress along the expanded tube was highest at the. transition region, and the residual stress in circumferential direction was found to be higher than the residual stress in axial direction. Hoop stress due to an internal pressure between primary and secondary side was analyzed to be 76 MPa and thermal stress was 45 MPa.

• Proceedings of the KSME Conference
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• 2000.11a
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• pp.837-842
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• 2000

In the coiling process of helical steam generator tubes of integral reactor SMART, a considerable amount of spring back, which induces dimensional inaccuracy and difficulty in fabrication, has been arised. In this research, an analytical model was derived to evaluate the amount of the spring back for steam generator tubes. The model was developed on the basis of beam theory and elastic-perfectly plastic material property. This model was extended to consider the effect of plastic hardening and the effect of the tensile force on the spring back phenomena. Parametric studies were performed for various design variables of steam generator tubes in order to minimize the spring back in the design stage. A sensitivity analysis has shown that the low yield strength, the high elastic modulus, the small helix diameter, and the large tube diameter result in a small amount of the spring back. The amount of the spring back can be controlled by the selection of adequate design values in the basic design stage and reduced to an allowable limit by the application of the tensile force to the tube during the coiling process.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.7 no.4
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• pp.19-25
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• 2011

Steam Generator(SG) tube is an important component of Nuclear Power Plant(NPP), which is the pressure boundary between the primary and secondary systems. The integrity of SG tube has been confirmed by the eddy current test every outage. The eddy current technique adopting bobbin probe is currently the primary technique for the steam generator tubing integrity assesment. The bobbin probe is one of the essential components which consist of the whole ECT examination system and provides us a decisive data for the evaluation of tube integrity. Until now, all of the ECT bobbin probes in Korea which is necessary to carry out inspection are imported from overseas. However, KHNP has recently developed the bobbin probe design technology and transferred it to domestic manufacturers to fabricate the probes. This study has been conducted to establish technical requirements applicable to the steam generator tube inspection using the bobbin probes fabricated by the domestic manufactures. The results have been compared with the results obtained by using foreign probe to identify the availability to the steam generator tube inspection. As a result, it is confirmed that the domestic bobbin probe is generally applicable to SG tube inspection in the NPPs.

• Journal of the Korean Society for Aeronautical & Space Sciences
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• v.35 no.3
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• pp.248-252
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• 2007

This article is related to the possibility for continuous operation of a liquid rocket engine when a portion of cryogenic propellant in the pipeline is vaporized. As a result of experimental studies imitating the formation of vapors in the flow, we confirmed the possibility of full gas-phase condensation in case temperature of cryogenic liquid is lower than it's saturation temperature in the pipeline. Empirical equation allowing to calculate a nonequilibrium condensation region in the steady flow of cryogenic liquid was obtained as a non-dimensional form and the fields of practical application were suggested.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05b
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• pp.473-480
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• 1996

영광 3/4호기 mid-loop 운전중 잔열제거(RHR) 기능 상실사고시 열수력적 현상을 최적 전산코드인 CATHARE2를 이용하여 해석하였다. 이러한 사고시 열수력적 현상은 일,이차측 냉각재 방출유로와 계통내 비응축성 가스의 거동에 의해 크게 영향을 받는다. 본 연구에서는 2개의 경우를 모의하였는데, 하나는 계통내 방출유로가 있는 경우이며 다른 하나는 방출유로가 없는 경우를 계산하였다. 이 때 사용된 가정은 다음과 같다. (가) 계통은 부분충수 운전 상태로 상부에 비응축성 가스나 증기로 가득 차 있다. (나) 증기발생기는 1대만이 이용 가능하고 이차측은 습식보관 상태이며, 보조급수는 공급되지 않고 이차측 압력은 대기압 상태이다 (다) 사고는 원자로 정지후 2일후 발생한다. 이와같은 조건하에서 사고시 계통 최대압력은 방출유로가 있는 경우 사고후 6,000 초에 0.27 MPa이며, 방출유로를 통한 유량은 총 2.4 kg/s이다. 이 방출유량을 외삽하여 계통수위가 고온관 바닦까지 도달하는데 걸린 시간은 사고후 약 5.67시간이다. 증기발생기 U-튜브를 통한 열전달에 의해 이차측 증기 발생으로 이차측 수위가 하락하면 증기발생기 reflux cooling은 제한을 받을 수 있다. 이 경우 이차측 수위가 U-튜브의 active 영역 상부까지 도달하는데 걸리는 시간은 사고후 약 10시간으로 계산되었다. 그러므로 이 경우 보조급수 공급 여유시간보다 노심 노출시간이 더 빨리 도달하여 노심을 손상시킨다. 사고시 수위지시계는 계통감압에 큰 영향을 주지 못하기 때문에 가능한 빨리 닫아 계통 inventory를 유지하는 것이 이차측 보조급수공급보다 우선한다.합한 설계방안으로 분석되었다.크다는 단점이 있다.TEX>$_2$O$_3$ 흡착제 제조시 TiO$_2$ 함량에 따른 Co$^{2＋}$ 흡착량과 25$0^{\circ}C$의 고온에서 ZrO$_2$와 $Al_2$O$_3$의 표면에 생성된 코발트 화합물을 XPS와 EPMA로 부터 확인하였다.인을 명시적으로 설명할 수 있다. 둘째, 오류의 시발점을 정확히 포착하여 동기가 분명한 수정대책을 강구할 수 있다. 셋째, 음운 과 정의 분석 모델은 새로운 언어 학습시에 관련된 언어 상호간의 구조적 마찰을 설명해 줄 수 있다. 넷째, 불규칙적이며 종잡기 힘들고 단편적인 것으로만 보이던 중간언어도 일정한 체계 속에서 변화한다는 사실을 알 수 있다. 다섯째, 종전의 오류 분석에서는 지나치게 모국어의 영향만 강조하고 다른 요인들에 대해서는 다분히 추상적인 언급으로 끝났지만 이 분석을 통 해서 배경어, 목표어, 특히 중간규칙의 역할이 괄목할 만한 것임을 가시적으로 관찰할 수 있 다. 이와 같은 오류분석 방법은 학습자의 모국어 및 관련 외국어의 음운규칙만 알면 어느 학습대상 외국어에라도 적용할 수 있는 보편성을 지니는 것으로 사료된다.없다. 그렇다면 겹의문사를 [-wh]의리를 지 닌 의문사의 병렬로 분석할 수 없다. 예를 들어 누구누구를 [주구-이-ν가] [누구누구-이- ν가]로부터 생성되었다고 볼 수 없다. 그러므로 [-wh] 겹의문사는 복수 의미를 지닐 수 없 다. 그러면 단수 의미는 어떻게 생성되는가\

• Journal of Radiation Protection and Research
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• v.14 no.2
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• pp.1-9
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• 1989

An analytic method for quantitative comparisions between the alternatives for radiation protection optimization is required to aid the decision making process. This paper introduces the dynamic Markov model to evaluate the effect of inservice inspection, testing, and repair activities of the plant on radiation protection. In the example to put the Markov model into practice, the steam generator inspection intervals which minimize expected cost and total exposure dose were determined using the data for Kori-2 unit and foreign plants. The results show that the effect of the radiation exposure on the steam generator inspection interval is determined by the cost rather than the radiation exposure. The Markov model used in the example can be applied easily to the domestic NPPs by replenishing the data and also can be used in evaluating the comparative priority between various alternatives for radiation protection optimization.

• Journal of Advanced Marine Engineering and Technology
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• v.33 no.8
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• pp.1144-1151
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• 2009

Refrigeration systems can be incorporated with non-adiabatic capillary tubes to improve their efficiency. The non-adiabatic capillary tube is constructed by joining the capillary tube with suction pipe to allow heat transfer between them, which is called capillary tube-suction line heat exchanger(SLHX). There are various joining methods and they may influence the characteristics of the refrigeration cycle. The present work aims to analyze the effect of widely-used two joining methods on the refrigeration cycle. The results show that soldered SLHX has much less thermal resistance than tapered SLHX but slightly outperforms in terms of coefficient of performance(COP) and cooling capacity. The soldered SLHX increased COP and cooling capacity of a refrigerator by 5.09% and 14.77% while the tapered SLHX did by 5.05% and 14.75%, respectively.