• Structural Engineering and Mechanics
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• 제5권2호
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• pp.149-165
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• 1997

A dynamic time-history analysis of the coupled internals and core in the vertical direction is performed as a part of the fuel assembly qualification program. To reflect the interaction between the fuel rods and grid cage, friction element is developed and is implemented. Also derived here is a method to calculate a hydraulic force on the reactor internals due to pipe break. Peak responses are obtained for the excitations induced from earthquake and pipe break. The dynamic responses such as fuel assembly axial forces and lift-off characteristics are investigated.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2001년도 춘계학술대회논문집A
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• pp.357-362
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• 2001

The pipe fracture tests were performed on 102mm-Sch.80 carbon steel pipe with various local wall thinning shapes, in order to understand failure behavior of thinned pipe. Pipe specimens were subjected to monotonic bending moment, using 4-points loading system, under internally pressurized condition. From the results of experiment, the failure mode, load carrying capacity, and deformability of local wall thinning pipe were investigated. Failure mode of thinned pipe depended on magnitude of internal pressure and thinning length as well as loading direction and thinning depth and angle. The variation in load carrying capacity and deformability of thinned pipe with length of thinned area was determined by stress type appled to thinning region and circumferential thinning angle. Also, the effect of internal pressure on failure behavior was dependent on failure mode of thinned pipe, and it promoted crack occurrence and mitigated local buckling at thinned area.

• 제어로봇시스템학회:학술대회논문집
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• 제어로봇시스템학회 2005년도 ICCAS
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• pp.346-351
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• 2005

The outlet feeder pipe thinning in a PHWR (Pressurized Heavy Water Reactor) is caused by high pressure steam flow inside the pipe, which is a well known degradation mechanism called FAC (Flow Assisted Corrosion). In order to monitor the degradation, the thickness of the outlet bends closed to the exit of the pressure tube should be measured and analyzed at every official overhaul. This paper develops a mobile feeder pipe inspection robot that can minimize the irradiation dose of human workers by automating the measurement process. The robot can move by itself on the feeder pipe by using an inch worm mechanism, which is constructed by two gripper bodies that can fix the robot body on the pipe, one extendable and contractable actuator, and a rotation actuator connected the two gripper bodies to move forward and backward, and to rotate in the circumferential direction

• International Journal of Control, Automation, and Systems
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• 제4권1호
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• pp.87-95
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• 2006

The outlet feeder pipe thinning in a PHWR (Pressurized Heavy Water Reactor) is caused by a high pressure steam flow inside the pipe, which is a well known degradation mechanism called a FAC (Flow Assisted Corrosion). In order to monitor the degradation, the thickness of the outlet bends close to the exit of the pressure tube should be measured and analyzed at every official overhaul. This paper describes a mobile feeder pipe inspection robot that can minimize the irradiation dose to human workers by automating the measurement process. The robot can move by itself on the feeder pipe by using an inch worm mechanism, which is constructed by two gripper bodies that can fix the robot body on to the pipe, one extendable and contractible actuator, and a rotation actuator connected to the two gripper bodies to move forward and backward, and to rotate in a circumferential direction.

• Nuclear Engineering and Technology
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• 제27권5호
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• pp.710-720
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• 1995

원자로에서 펌프에 의해 야기되는 맥동 압력은 원자로 내부 구조물에 진동과 손상을 줄 수 있기 때문에 관심이 증가되고 있다. 본 연구에서는 냉각관과 환형관(원자로 압력 용기와 노심 보호 지지대 사이)으로 구성된 기하 형태에서 펌프에 의해 야기되는 맥동 압력을 해석할 수 있는 수력학적 모델을 개발하였다. 수학적 지배 방정식은 압축성, 비점성 유체에 대해 선형화된 Navier-Stokes 방정식이다. 냉각관과 환형관을 따로 분리하여 해석하고 두영역의 커플링 영향을 고려하였다. 또한 본 기하 형태에서 펌프맥동 압력에 영향을 미치는 주요 기하 인자에 대한 평가를 수행하였다. 본 해석 결과와 실험차를 비교하여 만족할 만한 결과를 얻었다.

• Nuclear Engineering and Technology
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• 제55권8호
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• pp.3088-3101
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• 2023

A pressurized vessel-pipe-safety valve (PVPSV) combination is a commonly used configuration in nuclear power plants, and a good numerical model is essential for the system design, sizing and performance optimization. However, owing to the large-scale and cross-scale features, it is still a challenge to build a system level numerical model with both high accuracy and efficiency. To overcome this, a novel system level modeling method which can synthesize the advantages of various models is proposed in this paper. For system modeling, the analytical approach, the method of characteristics (MOC) and the surrogate model approach are respectively adopted to predict the dynamics of the pressure vessel, the connecting pipe and the safety valve, and different models are connected through data interfaces. With this system model, dynamic simulations were carried out and both the stable and the unstable system responses were obtained. For the model verification purpose, the simulation results were compared with those obtained from experiments and full CFD simulations. A good agreement and a better efficiency were obtained, verifying the ability of the model and the feasibility of the modeling method proposed in this paper.

• Geomechanics and Engineering
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• 제16권4호
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• pp.423-433
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• 2018

A three-dimensional model is constructed to simulate water infiltration in an unsaturated slope from a leaking pipe. Adaptive mesh refinement and time stepping are used, assisted by an automatic procedure for progressive steepening of the hydraulic property function for better convergence. The model is justified by comparing the simulated results with experimental data. Steady-state flow is investigated considering various pipe water pressures, locations and sizes of the opening, and soil layering. The opening size significantly affects the soaked zone around the pipe. Preferential flow dominates along the pipe longitudinal direction in the presence of a loose backfill around the pipe.

• Nuclear Engineering and Technology
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• 제32권1호
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• pp.67-76
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• 2000

Significant wall thinning due to flow accelerated corrosion(FAC）was recently reported in main feedwater pipes in 3 Korean pressurized water reactor(PWR) plants. The main feedwater pipes in one plant were repaired using overlay weld method at the outside of pipe, while those in 2 other plants were replaced with new pipes. In this study, the effect of the wall thinning in the main feedwater pipes on piping integrity was evaluated using finite element method. Especially, the effects of both the overlay weld repair and the stress concentration in notch-type thinned area on the piping integrity were investigated. The results are as follows : (1) The piping load carrying capacity may significantly decrease due to FAC. In special, the load carrying capacity of the main feedwater pipe was reduced by about 40% during about 140 months operation in Korean PWR plants. (2) By performing overlay weld repair at the outside of pipe, the piping load carrying capacity can increase and the stress concentration level in the thinned area can be reduced.

• 유체기계공업학회:학술대회논문집
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• 유체기계공업학회 2006년 제4회 한국유체공학학술대회 논문집
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• pp.551-554
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• 2006

Inhibition of propellant temperature rising in liquid propulsion rocket using cryogenic fluid as a propellant is very important. Especially propellant temperature rising during stand-by after filling and pre-pressurization can bring into cavitation in turbo-pump. One of the method preventing propellant temperature rising in cryogenic feeding system is recirculating propellant through the loop composed of propellant tank, feed pipe, and recirculation pipe. The circulation of propellant is promoted through gas-lift effect by gas injection to lower position of recirculation pipe. In this experiment liquid oxygen and gas helium is used as propellant and injection gas. Under atmospheric and pressurized tank ullage condition, helium injection flow-rate is varied to observe the variation of recirculating flow-rate and propellant temperature in the feed pipe. There is appropriate helium injection flow-rate for gas-lift recirculation system.

• Nuclear Engineering and Technology
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• 제54권8호
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• pp.3166-3175
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• 2022

The most numerical investigations of the thermal-hydraulic phenomena following the loss of the residual heat removal capability during the mid-loop operation of the pressurized water reactor were performed according to simplifications and are not sufficiently accurate. To perform more accurate and more reliable predictions of thermal-hydraulic accidents in a nuclear power plant using computational fluid dynamics codes, a more detailed methodology is needed. Modelling results identified that thermal stratification and natural convection are observed. Temperatures of lower monitoring points remain low, while temperatures of upper monitoring points increase over time. The water in the heated region, in the upper unheated region and the pipe region was well mixed due to natural convection, meanwhile, there is no natural convection in the lower unheated region. Water temperature in the pipe region increased after a certain time delay due to circulation of flow induced by natural convection in the heated and upper unheated regions. The modelling results correspond to the experimental data. The developed computational fluid dynamics methodology could be applied for modelling of two-component single/two-phase natural convection and thermal stratification phenomena during the mid-loop operation of the pressurized water reactor or other nuclear and non-nuclear installations at similar conditions.