• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 1994.04a
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• pp.116-121
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• 1994

지금까지 대형 구조물에 널리 사용되고 있는 Base Isolator는 특히, 지진하중에 대하여 상부 구조물의 지반분리(base isolation)을 이용한 지진제어를 목적으로 하기 때문에 Seismic Isolator라 한다. 일반적으로 면진베어링에는 LRB(Laminated Rubber Bearing) 방식과 R-FBI(resilient-friction base isolator) 방식으로 크게 나눌 수 있다. LRB 방식은 가장 널리 사용되는 면진베어링으로써 방진고무를 주된 재료로 하고 수직강성을 보강하기 위하여 steel plate를 조합하여 제작하며, 초기강성 및 에너지 소산능력을 증가시키기 위하여 단면중앙에 납(lead plug)를 삽입하기도 한다. R-FBI 방식은 방진고무 적층판 내부에 미끄럼판을 가지고 있으므로 LRB 방식에 비하여 더 큰 수평변위를 발생시킬 수 있다. 이번에 설계 제작한 면진베어링은 LNG 저장탱크의 Seismic Isolation을 목적으로 적용대상의 사양에 맞추어 설계 제작하고 성능평가 시험을 수행하여 LNG 저장탱크, 원자로, 대형 건축물등 지진으로부터 보호되어야 하는 대형구조물의 방진재로 적용할 수 있는가를 평가하고자 한다. 본 논문에서는 제작된 면진베어링의 설계를 검증하고 방진고무(HDR)재료의 물리적 특성시험, 축소모델에 의한 정적, 동적시험을 통하여 시험방법을 소개하고 이러한 시험결과를 기초로 하여 면진베어리의 성능을 평가하였으며, 면진베어링의 온도변화, 외부 수직하중의 변화등에 따른 특성변화와 LNG 저장탱크와 면진베어링의 체결방법에 따른 시험으로 체결방법을 검증하였으며, 대상물의 사양에 적합한가를 고찰하였다.

• Transactions of the Korean Society for Noise and Vibration Engineering
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• v.25 no.1
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• pp.48-57
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• 2015

In a sodium-cooled fast reactor, which is a Generation-IV reactor, refueling is conducted by rotating, but not opening, the reactor head to prevent a reaction between the sodium, water and air. Therefore, an inspection technique that checks for the presence of any obstacles between the reactor core and the upper internal structure, which could disturb the rotation of the reactor head, is essential prior to the refueling of a sodium-cooled fast reactor. To this end, an ultrasound-based inspection technique should be employed because the opacity of the sodium prevents conventional optical inspection techniques from being applied to the monitoring of obstacles. In this study, a ranging inspection technique using a plate-type ultrasonic waveguide sensor was developed to monitor the presence of any obstacles between the reactor core and the upper internal structure in the opaque sodium. Because the waveguide sensor installs an ultrasonic transducer in a relatively cold region and transmits the ultrasonic waves into the hot radioactive liquid sodium through a long waveguide, it offers better reliability and is less susceptible to thermal or radiation damage. A 10 m horizontal beam waveguide sensor capable of radiating an ultrasonic wave horizontally was developed, and beam profile measurements and basic experiments were carried out to investigate the characteristics of the developed sensor. The beam width and propagation distance of the ultrasonic wave radiated from the sensor were assessed based on the experimental results. Finally, a feasibility test using cylindrical targets (corresponding to the shape of possible obstacles) was also conducted to evaluate the applicability of the developed ranging inspection technique to actual applications.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.18 no.2_spc
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• pp.291-303
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• 2020

The Kori-Unit 1 nuclear power plant, which is scheduled to be decommissioned after permanent shutdown, is expected to generate large amounts of various types of radioactive waste during the decommissioning process. Among these, nuclear reactors and internal structures have high levels of radioactivity and the dismantled structure must have the proper size and weight on the primary side. During decommissioning, it is important to prepare an appropriate and efficient disposal method through analysis of the disposal status and the legal restrictions on wastes generated from the reactors and internal structures. Nuclear reactors and internal structures generate radioactive wastes of various levels, such as medium, very low, and clearance. A radiation evaluation indicates that wastes in the clearance level are generated in the reactor head and upper head insulation. In this study, a clearance waste safety evaluation was conducted using the RESRAD-RECYCLE code, which is a safety evaluation code, based on the activation evaluation results for the clearance level wastes. The clearance scenario for the target radioactive waste was selected and the maximum individual and collective exposure doses at the time of clearance were calculated to determine whether the clearance criteria limit prescribed by the Nuclear Safety Act was satisfied. The evaluation results indicated that the doses were significantly low, and the clearance criteria were satisfied. Based on the safety assessment results, an appropriate metal recycle and disposal method were suggested for clearance, which are the subject of the deregulation of internal structures of nuclear power plant.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2012.04a
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• pp.734-738
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• 2012

A comprehensive vibration assessment program for the Advanced Power Reactor 1400 reactor vessel internals is established in accordance with the United States Nuclear Regulatory Commission Regulatory Guide 1.20 Revision 3. This paper is related to instruments and measurement locations based on the vibration and stress response analysis results at the inner barrel assembly top plate in the reactor. The analysis results of the inner barrel assembly top plate in the reactor show that the deterministic stress and deformation due to the reactor coolant pump induced pressure pulsations are larger than the random stress and deformation induced by the flow turbulence. The selection of the instruments and measurement locations at Inner barrel assembly top plate in the reactor is essential requirements and very important study process for the vibration and stress measurement program in comprehensive vibration assessment program for the Advanced Power Reactor 1400 reactor vessel internals.

• Proceedings of the KSME Conference
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• 2004.11a
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• pp.435-440
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• 2004

The nonlinear impacts between the Control Element Drive Mechanisms (CEDMs) seismic cap plates installed on the CEDM top of a pressurized water reactor are studied with the dynamically reduced models of the CEDM and Integrated Head Assembly (IHA). It is important to develope nonlinear models considering the gap effects between the plates. In order to simulate impacts, reduced models for the primary structures, such as CEDM and IHA, are developed through simplifying detailed models, and the nonlinear structural analysis is performed under seismic loading conditions. The responses are examined in various gap sizes depending on the reactor operating conditions.

• Transactions of the Korean Society for Noise and Vibration Engineering
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• v.22 no.5
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• pp.474-479
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• 2012

A comprehensive vibration assessment program for the advanced power reactor 1400(APR1400) reactor vessel internals is established in accordance with the united states nuclear regulatory commission regulatory guide 1.20 revision 3. This paper is related to instruments and measurement locations based on the vibration and stress response analysis results of the inner barrel assembly top plate in APR1400. The analysis results of the inner barrel assembly top plate in the reactor show that the deterministic stress and deformation due to the reactor coolant pump induced pressure pulsations are larger than the random stress and deformation induced by the flow turbulence. The selection of the instruments and measurement locations at inner barrel assembly top plate in the reactor is essential requirements and very important study process for the vibration and stress measurement program in comprehensive vibration assessment program for APR1400 reactor vessel internals.

• Journal of Radiation Protection and Research
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• v.39 no.1
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• pp.21-29
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• 2014

A new cold neutron triple-axis spectrometer (Cold-TAS) was recently constructed at the 30 MWth research reactor, HANARO. The spectrometer, which is composed of neutron optical components and radiation shield, required a redesign of the segmented monochromator shield due to the lack of adequate support of its weight. To shed some weight, lowering the height of the segmented shield was suggested while adding more radiation shield to the top cover of the monochromator chamber. To investigate the radiological effect of such change, we performed MCNPX simulations of a few different configurations of the Cold-TAS monochromator shield and obtained neutron and photon intensities at 5 reference points just outside the shield. Reducing the 35% of the height of the segmented shield and locating lead 10 cm from the bottom of the top cover made of polyethylene was shown to perform just as well as the original configuration as radiation shield excepting gamma flux at two points. Using gamma map by MCNPX, it was checked that is distribution of gamma. Increased flux had direction to the top and it had longer distance from top of segmented shield. However, because of reducing the 35% of the height, height of dissipated gamma was lower than original geometry. Reducing the 35% of the height of the segmented shield and locating lead 10cm from the bottom of the top cover was selected. After changing geometry, radiation dose was measured by TLD for confirming tester's safety at any condition. Neutron(0.21 ${\mu}Svhr^{-1}$) and gamma(3.69 ${\mu}Svhr^{-1}$) radiation dose were satisfied standard(6.25 ${\mu}Svhr^{-1}$).

• Journal of the Korean Society for Nondestructive Testing
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• v.26 no.5
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• pp.297-305
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• 2006

The thermal ratcheting deformation at the reactor baffle and upper internal structure of the liquid metal reactor (LMR) can occur due to movement of the hot sodium free surface. In in-service inspection of reactor internals of LMR, a new inspection technique should be developed for the detection of the thermal ratcheting damage. In this study, an inspection technique using ultrasonic guided wave is proposed for the detection of the thermal ratcheting damage of cylindrical vessels. A 316L stainless steel cylindrical shell specimen has been prepared. The thermal ratchet structural tests were cyclically performed by heat-up up to $550^{\circ}C$ with steep temperature gradients along the axial direction after cool-down by cooling water. Ultrasonic guided wave propagation has been characterized by analysis of dispersion curve of the stainless steel plate. The zero-order antisymmetric $A_0$ guided wave has been selected as the optimal mode for detection of the ratcheting deformation. It is confirmed that the thermal ratcheting deformation can be detected by the measurement of transit time difference of circumferentially propagated $A_0$ guided waves.