• 한국유체기계학회 논문집
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• 제17권6호
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• pp.109-114
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• 2014

After the Fukushima accident in 2011, it was revealed that nuclear power plant has the vulnerability to SBO accident and its extension situation without sufficient cooling of reactor core resulting core meltdown and radioactive material release even after reactor shutdown. Many safety systems had been developed like PAFS, hybrid SIT, and relocation of RPV and IRWST as a part of steps for the Fukushima accident, however, their applications have limitation in the situation that supply of feedwater into reactor is impossible due to high pressure inside reactor pressure vessel. The concept of hybrid heat pipe with control rod is introduced for breaking through the limitation. Hybrid heat pipe with control rod is the passive decay heat removal system in core, which has the abilities of reactor shutdown as control rod as well as decay heat removal as heat pipe. For evaluating the cooling performance hybrid heat pipe, a commercial CFD code, ANSYS-CFX was used. First, for validating CFD results, numerical results and experimental results with same geometry and fluid conditions were compared to a tube type heat pipe resulting in a resonable agreement between them. After that, wall temperature and thermal resistances of 2 design concepts of hybrid heat pipe were analyzed about various heat inputs. For unit length, hybrid heat pipe with a tube type of $B_4C$ pellet has a decreasing tendency of thermal resistance, on the other hand, hybrid heat pipe with an annular type $B_4C$ pellet has an increasing tendency as heat input increases.

• 대한용접접합학회:학술대회논문집
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• 대한용접접합학회 1999년도 특별강연 및 춘계학술발표대회 개요집
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• pp.28-31
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• 1999

• Nuclear Engineering and Technology
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• 제28권4호
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• pp.405-414
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• 1996

In this paper, the unsteady 2-dimensional turbulent flow model for thermal stratification in a pressurizer surge line of PWR plant is proposed to numerically investigate the heat transfer and flow characteristics. The turbulence model is adapted to the low Reynolds number K-$\varepsilon$ model (Davidson model). The dimensionless governing equations are solved by using the SIMPLE (Semi-Implicit Method for Pressure Linked Equations) algorithm. The results are compared with simulated experimental results of TEMR Test. The time-dependent temperature profiles in the fluid and pipe nil are shown with the thermal stratification occurring in the horizontal section of the pipe. The corresponding thermal stresses are also presented. The numerical result for thermal stratification by the outsurge during heatup operation of PWR shows that the maximum dimensionless temperature difference is about 0.83 between hot and cold sections of pipe well and the maximum thermal stress is calculated about 322MPa at the dimensionless time 28.5 under given conditions.

• 대한기계학회논문집A
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• 제29권5호
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• pp.734-742
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• 2005

The present study proposes the integrity evaluation model for a straight pipe with local wall thinning defect, which reflects the characteristics of training shape and loading condition in the Piping of nuclear power plant. For this purpose, a series of finite element analyses are performed under various defect geometries and loading conditions, and real pipe experiment data performed previously is employed. The model includes the effect of thinning length as well as thinning depth and width, and also it considers the combined loading effect between internal pressure and bending moment. The proposed model has been validated using the results of finite element analysis and pipe experiment data. The results indicate that the proposed model provides more reliable predictions of pipe failure than the current existing model, in terms of accuracy, consistency, and conservativeness of results.

• 한국압력기기공학회 논문집
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• 제19권1호
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• pp.52-59
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• 2023

The welding part of the valve needs immediate action when leakage occurs due to cracks or damage. In order to repair leakage of the welding part, the valve must be separated from the pipe or replaced with a new valve. However, it is difficult to remove the valve while operating the power plant. This study presents a method to remove leakage by precisely processing the gap between the clamp and the incision part within 0.1mm while installed in the pipe system. If the external leakage is removed using a clamp on the welding part without removing the valve during operation, the time and cost required for maintenance can be reduced.

• 한국압력기기공학회 논문집
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• 제6권1호
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• pp.78-82
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• 2010

The feeder pipe of Main Primary Heat Transfer System in Wolsong Nuclear Power Plant was inspected by the Ultrasonic Phase Array technique in 2010. It is the first time to apply this method to the construction at Nuclear Power Plant in Korea. The time required for UT technique is less than RT method. The UT method doesn't need to evacuate personnel who works nearby inspecting area and doesn't need to wait developing of film. For these reasons, the UT method is the fastest method among the volumetric inspections. As a result of the examination, it became clear that main defect of the feeder pipe is the Lack of fusion in the welded area. Moreover, the rate of defect was reduced gradually as improvement of welder's skill. If welding machine has problem, the defect has tended to same pattern(occurred same position in the welding area) but these defects were founded without specific rules. For these reasons, the creation of defect is dependent on the skill of worker not on the automatic welding machine. This evaluation of defect signal and collecting data would be useful to further examination in ISI.

• Corrosion Science and Technology
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• 제17권6호
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• pp.292-300
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• 2018

The modeling of 3D finite elements based on CAD data has been used to detect sites of corrosion defects in buried pipes. The results generated sophisticated profiles of electrolytic potential and vectors of current distributions on the earth surface. To identify the location of defects in buried pipes, the current distribution on the earth surface was projected to a plane of incidence that was identical to the pipe locations. The locations of minimum electrolytic potential value were found. The results show adequate match between the locations of real and expected defects based on modeling. In addition, the defect size can be calculated by integrating the current density curve. The results show that the defect sizes were $0.74m^2$ and $0.69m^2$, respectively. This technology may represent a breakthrough in the detection of indirect damage in various cases involving multiple defects in size and shape, complex/cross pipe systems, multiple anodes and stray current.

• 대한전기학회:학술대회논문집
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• 대한전기학회 2004년도 학술대회 논문집 정보 및 제어부문
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• pp.635-637
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• 2004

The containment of the nuclear power plant is the last barrier of radiation release when the reactor coolant pipe rupture is occurred. Each plant has to be tested every 5 years whether the containment leak rate meets its technical specifications. We have developed the leak rate test system and in this paper, we describe the results of the uncertainty analysis on the measurement channels and its propagation to the calculation results.

• 비파괴검사학회지
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• 제30권4호
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• pp.295-301
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• 2010

원자력발전소 탄소강 배관의 유체가속부식은 주요 경년열화 현상이며 발전소의 성능 및 안전성을 저해할 수 있다. 유체가속부식 검사는 보온재 제거 및 설치로 상당한 비용이 수반되므로 최근에 보온재 제거가 필요 없고 원거리 검사가 가능한 유도초음파에 대한 관심이 점점 증가되고 있다. 유체가속부식 검출에 유도초음파 적용이 가능하다면 검사 비용 절감이 예상된다. 본 연구의 목적은 유체가속부식 손상 유무를 확인하고 결함 검출능을 결정하기 위함이다. 본 연구에서는, 실제 유체가속부식 손상 시험편의 엘보우 첫 번째 용접부와 두 번째 용접부의 진폭 감쇄비를 측정하기 위하여 3가지 검사 기법을 사용하였다. 연구 결과, 유체가속부식 손상을 검출하기 위한 최적의 검사 기법과 최소 결함 검출능을 도출하였다.

• Nuclear Engineering and Technology
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• 제54권10호
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• pp.3745-3765
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• 2022

The concrete structures related to nuclear safety are threatened by accidental impact loadings, mainly including the low-velocity drop-weight impact (e.g., spent fuel cask and assembly, etc. with the velocity less than 20 m/s) and high-speed projectile impact (e.g., steel pipe, valve, turbine bucket, etc. with the velocity higher than 20 m/s), while the existing studies are still limited in the impact resistant design of nuclear power plant (NPP), especially the primary RC slab. This paper aims to propose the numerical simulation and theoretical approaches to assist the impact-resistant design of RC slab in NPP. Firstly, the continuous surface cap (CSC) model parameters for concrete with the compressive strength of 20-70 MPa are fully calibrated and verified, and the refined numerical simulation approach is proposed. Secondly, the two-degree freedom (TDOF) model with considering the mutual effect of flexural and shear resistance of RC slab are developed. Furthermore, based on the low-velocity drop hammer tests and high-speed soft/hard projectile impact tests on RC slabs, the adopted numerical simulation and TDOF model approaches are fully validated by the flexural and punching shear damage, deflection, and impact force time-histories of RC slabs. Finally, as for the two low-velocity impact scenarios, the design procedure of RC slab based on TDOF model is validated and recommended. Meanwhile, as for the four actual high-speed impact scenarios, the impact-resistant design specification in Chinese code NB/T 20012-2019 is evaluated, the over conservation of which is found, and the proposed numerical approach is recommended. The present work could beneficially guide the impact-resistant design and safety assessment of NPPs against the accidental impact loadings.