• 한국콘크리트학회:학술대회논문집
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• 한국콘크리트학회 1994년도 가을 학술발표회 논문집
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• pp.285-290
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• 1994

The influence of elevated temperatures on the mechanical properties of concrete is important for fire-resistance studies and also for understanding the behavior of containment vessel, such as nuclear reactor pressure vessels, during service and ultimate condition. The present study is to clarify the damage/deterioration of concrete structures that are subjected to high temperature exposure. To this end, comprehensive experiments are conducted. The major test variables are the peak temperatures, rate of temperature increase, and sustained duration at peak temperature. The results include weight loss residual compressive strength and stress-strain curve. From those results, residua compressive strength formula and stress-strain relationship are proposed.

• 비파괴검사학회지
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• 제34권3호
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• pp.254-259
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• 2014

중수로 원자로는 한 개의 원자로용기로 구성된 경수로와는 달리 약 380여개의 연료채널(fuel channel)로 구성되어 있다. 연료채널을 구성하는 압력유지 기기인 압력관(pressure tube)은 지르코니움 합금(Zr-2.5wt% Nb) 재질로서 치수는 내경이 103.4 mm, 두께가 약 4.19 mm, 길이가 6.36 m인 튜브 형태의 관이다. 압력관은 내부에 핵연료 다발과 냉각재가 내장되며 압력관의 기능은 연료를 지지하고 열수송 유체인 중수($D_2O$)를 이송한다. 압력관의 단순한 기하학적인 형상으로 인하여 자동화 비파괴검사가 가능하고 접근성이 우수하다. 연료채널은 경수로형 원전과 동일하게 설치전과 운전중에 원자력안전위원회 법령 요건에 따라 주기적으로 엄격한 비파괴검사를 수행하여 건전성을 확인한다. 연료채널의 주기적 비파괴검사에는 초음파탐상 및 와전류탐상검사 기법을 적용한 체적 비파괴검사 기술이 적용된다. 이중에서 와전류탐상검사 기법은 초음파탐상검사에서 검출된 결함의 확인을 위한 보충검사기술로 적용되고 있지만 표면결함에 대한 검출능이 초음파탐상검사 기법보다 우수한 장점을 가지고 있다. 본 논문에서는 압력관 내부 표면 비파괴검사에 적용되고 있는 와전류탐상검사 기술의 압력관 내면에 발생할 수 있는 결함의 검출 및 깊이 측정 특성에 대한 연구결과를 기술하였다. 즉, 와전류검사 기술은 압력관 내면에 발생할 수 있는 아주 미세한 결함을 매우 우수한 분해능으로 검출할 수 있으므로 초음파탐상검사 결과 확인을 위한 보충기술로서 매우 유용하지만, 결함의 깊이 측정은 오차가 매우 크게 발생하므로 결함 깊이 측정에는 적합하지 않고 오직 표면결함 검출에만 적용하는 것이 바람직하다.

• 한국산학기술학회논문지
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• 제5권3호
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• pp.227-231
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• 2004

It is the common features of the integral reactors that the main components of the RCS are installed within the reactor vessel, and so there are no any flow pipes connecting the steam generator or the pump whose type is the axial flow. Due to no any flow pipes, it is impossible to measure the differential pressure at the RCS of the integral reactors, and it also makes impossible measure the flow-rate of the reactor coolant. As a alternative method, the method by the measurement of the pump power of the axial flow pump has been introduced in this study. Up to now, we did not found out a precedent which the pump power is used for the flow-rate calculation at normal operation of the commercial nuclear power plants. The objective of the study is to embody the flow-rate calculation method by the measurement of the pump power in an integral reactor. As a result of the study, we could theoretically reason that the capacity-head curve and capacity-shaft power curve around the rated capacity with the high specific-speeded axial flow pumps have each diagonally steep incline but show the similar shape. Also, we could confirm the above theoretical reasoning from the measured result of the pump motor inputs. So, it has been concluded that it is possible to calculate the flow-rate by the measurement of the pump motor inputs.

• Nuclear Engineering and Technology
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• 제26권2호
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• pp.257-264
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• 1994

원자력발전소 중대사고시 용융된 노심과 잔류냉각수가 증기폭발을 일으켜 원자로 격납용기의 건전성을 위협할 수 있다. 본 연구에서는 증기폭발을 모사할 수 있는 실험 장치를 제작하고, 물과 프레온을 사용하여 증기폭발실험을 수행하였다. 이때 고속카메라를 사용하여 폭발현상을 관측하였고, 동압측정기와 압력분출관을 이용하여 생성되는 폭발압력과 기계적인 에너지를 계측하였다. 이를 토대로 증기폭발의 중요인자들(물의 온도, 물의 주입속도, 물의 주입 시간, 그리고 냉매의 깊이)에 대한 민감도 분석을 수행하였다. 그리고, 압력용기 바닥의 구조물이 용융/냉각재의 반응에 미치는 영향을 살펴보기위하여 실험용기 내부에 그리드를 설치하여 폭발실험을 실시하였다. 물/프레온의 폭발실험에서 계측된 기계적에너지를 이용한 에너지효율은 0.5∼l.6%인 것으로 계산되었다.

• 한국염색가공학회지
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• 제33권3호
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• pp.141-146
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• 2021

Supercritical fluid has excellent dissolving power for various materials based on low viscosity and high diffusion coefficient and is used as solvents in various chemical processes. However, its industrial application can be very tricky because the design, especially the size of the supercritical apparatus, should be carefully chosen to optimize the cost and the production of supercritical fluidic state. And the first step of the supercritical fluid apparatus design is to choose the appropriate size of the reactor vessel to produce supercritical fluid above its critical pressure and temperature. Therefore, this study aims to analyze thermodynamic behaviors of dichloromethane based on ideal gas, van der Waals, Redlich-Kwong, Soave-Redlich-Kwong, and Peng-Robinson equations of state. The correlation between the volume and pressure of dichloromethane at 200℃ was revealed and it can be used to design the optimized size of the supercritical apparatus for industrial production.

• Corrosion Science and Technology
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• 제20권5호
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• pp.308-314
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• 2021

Neutron dose level at bottom head of a reactor pressure vessel (RPV) was calculated using reactor vessel neutron transport for a Korean nuclear power plant A. At 34 EFPY with a 40-year (2042) design life after plating repair, irradiation fast neutron effect was 6.6x10¹⁵ n/cm². As helium(He) gas can be generated by Ni only at 1/10⁶ level of 5 × 10²¹ n/cm², He generation possibility in the Ni plating layer is very little during 40 years of operation (2042, 34 EFPY). Thermal neutrons can significantly affect the generation of He from Ni metal. At 10 years after a repair, He can be generated at a level of about 0.06 appm, a level that can add general welding repair without any consideration. After 40 years of repair, 9.8 appm of He may be generated. Although this is a rather high value, it is within the range of 0.1 to 10 appm when welding repair can be applied. Clad repair by Ni electroplating technology is expected to greatly improve the operation efficiency by improving the safety and shortening the maintenance period of the nuclear power plant.

• 대한기계학회논문집A
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• 제26권3호
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• pp.445-451
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• 2002

The refractory-lined pipe is used to protect the system from high-temperature of the internal flow. The property of the refractory has an effect upon the stress analysis for fluid catalytic cracking(FCC) unit piping design. The equivalent elastic modulus and density considering steel and refractory must be applied in the stress analysis of the system. In the research, the theoretical method to obtain the value of the equivalent property is introduced and then the parametric analysis is carried out to understand the characteristic of the material properties, and the stress analysis is performed with reactor, the part of FCC unit.

• 한국콘크리트학회:학술대회논문집
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• 한국콘크리트학회 1994년도 가을 학술발표회 논문집
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• pp.291-300
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• 1994

The influence of elevated temperatures on the mechanical properties of concrete is important for fire-resistance studies and also for understanding the behavior of containment vessel, such as nuclear reactor pressure vessels, during service and ultimate condition. The present study is to clarify the damage/deterioration of concrete structures that are subjected to high temperature exposure. To this end, comprehensive experiments are conducted. The major test variables are the peak temperatures, rate of temperature increase, and sustained duration at peak temperature. The results include weight loss residual compressive strength and stress-strain curve. From those results, residua compressive strength formula and stress-strain relationship are proposed.

• 대한기계학회논문집A
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• 제26권1호
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• pp.105-112
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• 2002

In recent years, the integrity of reactor Pressure Vessel(RPV) under pressurized thermal shock (PTS) accident has been treated as one of the most critical issues. Under PTS condition, the combination of thermal and mechanical stress by steep temperature gradient and internal pressure causes considerably high tensile stress at the inside of RPV wall. As a result, cracks on inner surface of RPV may experience elastic-plastic behavior which can be characterized by J-integral. In such a case, however, J-integral may possibly lose its vapidity due to the constraint effect. The degree of constraint effect is influenced by the loading mode, crack geometry and material properties. In this paper, in order to investigate the effect of clad thickness and crack geometry on constraint effect, three dimensional finite element analyses were performed for various surface cracks. Total of 27 crack geometries were analyzed and results were presented by a two-parameter characterization based on the J-integral and the f-stress.

• Journal of Welding and Joining
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• 제22권5호
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• pp.38-45
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• 2004

The present study is to investigate the effect of heat input on the microstructure, tensile properties and toughness of single-pass submerged arc bead-in-groove welds produced on SA508 class 3 steels. The heat input was varied in the range of 1.6, 3.2 and 5.0 kJ/mm. The toughness of weld metals was evaluated by using subsize Charpy V-notch specimens in the temperature range of -19$0^{\circ}C$ to 2$0^{\circ}C$. The weld microstructure and fractography were observed by optical and scanning electron microscopies, respectively. With increasing heat inputs, tensile strength and hardness of weld metals were decreased while elongation was increased. The poor notch toughness at 1.6 kJ/mm was attributed to the formation of ferrite with aligned second phase and banitic microstructure with high yield strength while that at 5.0 kJ/mm was due to the presence of grain boundary and polygonal ferrites. The microstructure of the intermediate energy input welds consisted of a high proportion of acicular ferrite with limited polygonal ferrites, which provide improved notch toughness.