• Nuclear Engineering and Technology
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• 제53권9호
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• pp.3085-3099
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• 2021

Investigations of large commercial aircraft impact effect on nuclear power plant (NPP) buildings have been drawing extensive attentions, particularly after the 9/11 event, and this paper aims to numerically assess the damage and vibrations of NPP buildings subjected to aircrafts crash. In Part I of present paper, two shots of reduce-scaled model test of aircraft impact on NPP were conducted based on the large rocket sled loading test platform. In the present part, the numerical simulations of both scaled and prototype aircraft impact on NPP buildings are further performed by adopting the commercial program LS-DYNA. Firstly, the refined finite element (FE) models of both scaled aircraft and NPP models in Part I are established, and the model impact test is numerically simulated. The validities of the adopted numerical algorithm, constitutive model and the corresponding parameters are verified based on the experimental NPP model damages and accelerations. Then, the refined simulations of prototype A380 aircraft impact on a hypothetical NPP building are further carried out. It indicates that the NPP building can totally withstand the impact of A380 at a velocity of 150 m/s, while the accompanied intensive vibrations may still lead to different levels of damage on the nuclear related equipment. Referring to the guideline NEI07-13, a maximum acceleration contour is plotted and the shock damage propagation distances under aircraft impact are assessed, which indicates that the nuclear equipment located within 11.5 m from the impact point may endure malfunction. Finally, by respectively considering the rigid and deformable impacts mainly induced by aircraft engine and fuselage, an improved Riera function is proposed to predict the impact force of aircraft A380.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2005년도 춘계학술발표회
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• pp.587-588
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• 2005

• Nuclear Engineering and Technology
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• 제52권11호
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• pp.2638-2651
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• 2020

Nozzle corner cracks present at the intersection of reactor pressure vessels (RPVs) and inlet or outlet nozzles have been a persistent problem for a number of years. The fracture analysis of such nozzle corner cracks is very important and critical for the efficient design and assessment of the structural integrity of RPVs. This paper aims to perform an engineering critical assessment of RPVs with nozzle corner cracks subjected to several transients accompanied by pressurized thermal shocks. The critical crack size of the RPV model with nozzle corner cracks under transient loading is evaluated on failure assessment curve. In particular, the influence of cladding on the crack initiation of nozzle corner crack under thermal transients is studied. The influence of primary internal pressure and secondary thermal stress on the stress field at nozzle corner and SIF at crack front is analyzed. Finally, the influence of different crack size and crack shape on the final critical crack size is analyzed.

• 동력기계공학회지
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• 제20권2호
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• pp.66-72
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• 2016

As a method to improve the equipment reliability, early warning researches that can be detected fault symptom of an equipment at an early stage are being performed out among developed countries. In this paper, when abnormal signal is input to actual normal signal of a pump, early detection studies on pump's fault symptom were carried out with auto-associative kernel regression as an advanced pattern recognition algorithm. From analysis, correlations among power of motor driving pump, discharge flow of pump, power output of pump, and discharge pressure of pump are exited. When the abnormal signal is input to one of those normal signals, the other expected values are changed due to the influence of the abnormal signal. Therefore, the fault symptom of pump through the early-warning index is able to detect at an early stage.

• 한국압력기기공학회 논문집
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• 제15권1호
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• pp.1-7
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• 2019

In this paper, the regulations and technical standards related to qualification of safety grade equipment in nuclear power plants are critically reviewed with the qualification procedure in terms of structures, systems, and equipment in nuclear power plants. These facilities should be designed and constructed to protect from natural conditions or disasters and to perform their safety functions even in case of postulated accidents. Equipment Qualification is to demonstrate that the safety related equipment is designed and constructed to perform their safety functions under normal and accident conditions. It is classified into environmental qualification and seismic qualification.

• 한국지진공학회:학술대회논문집
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• 한국지진공학회 2000년도 추계 학술발표회 논문집 Proceedings of EESK Conference-Fall 2000
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• pp.215-223
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• 2000

In this study, the effect of stiffness ratio between base frame and anchorgae is evaluated and the seismic verification of nuclear power plant equipment anchorage is performed for typical equipment. The stiffness ratio between base frame and anchorage is mainly controlled by the effective height of side wall plate. And, the change of that stiffness ratio cause the large shift or ovreturning axis of equipment base. This shift of overturning axis of equipment base is able to reduce the factor of safety about 10%. Therefore, the adequate method for evaluating of effective height of side wall is required as further study.

• Nuclear Engineering and Technology
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• 제55권8호
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• pp.2905-2918
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• 2023

The valve assembly used in nuclear power plants is important safety-related equipment. In the new standard, the physical attributes are measured using a valve diagnosis test, which is used in the expansion to other non-tested valves using a quantitative test-basis methodology. With a motor-operated actuator, the state of stem's lubrication is related to physical attributes such as the stem factor and the friction coefficient. This study analyzed the numerical transient of fluid and solid lubrication with a squeeze film effect due to the loading rate on the stem and the stem nut using the experimental data. The differential equation that governs the motion mechanism of the stem and stem nut is established and analyzed. The flow rate, the fluid and the solid contact forces are calculated with the friction coefficient. Finally, we found that a change in the friction coefficient results from a change of the shear force in the solid contact mode during the interchange process between the solid contact mode and the fluid contact mode. The qualitative understanding of the squeeze film effect is expanded quantitatively for forces, thread surface distance, velocity, and acceleration, with consideration of the metal solid contact and fluid contact.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1978년도 제10회 학술 발표회 초록
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• pp.116.2-116
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• 1978

• Nuclear Engineering and Technology
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• 제37권2호
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• pp.177-184
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• 2005

The Multivariate State Estimation Technique (MSET) is being used in Nuclear Power Plants for sensor and equipment condition monitoring. This paper presents the use of regularization methods for optimizing MSET's predictive performance. The techniques are applied to a simulated data set and a data set obtained from a nuclear power plant currently implementing empirical, on-line, equipment condition monitoring techniques. The results show that regularization greatly enhances the predictive performance. Additionally, the selection of prototype vectors is investigated and a local modeling method is presented that can be applied when computational speed is desired.

• 신재생에너지
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• 제16권4호
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• pp.49-64
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• 2020

Photovoltaic and Thermal collector (PV/T) systems are renewable energy devices that can produce electricity and heat energy simultaneously using solar panels and heat exchangers. Since PV/T systems are exposed to the outdoors, their reliability is affected by various environmental factors. This paper presents a reliability test for a PV/T system and evaluates the test results. The reliability assessment entails performance, environment, safety, and life tests. The factor that had the greatest influence on the life of the system was the hydraulic pressure applied to the heat exchanger. A test was conducted by repeatedly applying pressure to the PV/T system, and a reliability analysis was conducted based on the test results. As a result, the shape parameter (β) value of 5.6658 and the B10life 308,577 cycles at the lower 95% confidence interval were obtained.