• Nuclear Engineering and Technology
• /
• 제55권9호
• /
• pp.3252-3259
• /
• 2023

The influence of hold-down spring (HDS) on the vibration characteristics of core barrel is studied in this paper. First, the vibration characteristics experiment of core barrel was carried out with four type of different hold-down spring. These hold-down springs represent the same hold-down force under different spring stiffness and different hold-down force under the same spring stiffness. And then a new finite element method for researching the influence of hold-down spring on the vibration characteristics of core barrel was presented. This new method could consider the influence of the hold-down force and the spring stiffness at the same time. The results suggest that, the hold-down force and friction have greater influence on the vibration characteristics of core barrel than the spring stiffness, and the influence is nonlinear. The influence of the boundary condition on beam mode is greater than that on shell mode for core barrel.

• 한국소음진동공학회:학술대회논문집
• /
• 한국소음진동공학회 1991년도 춘계학술대회논문집; 한국해사기술연구소, 대전; 1 Jun. 1991
• /
• pp.163-166
• /
• 1991

현재 PWR의 동력로의 값은 대당 10억불이나 되는데, 원래 20년 수명을 예 측하고 설계된 것이나, 이를 두 배로 늘려서 40년이 수명을 가질 수 있겠는 가 하는 문제가 크게 대두되었다. 본 연구는 가장 중요한 구조물인 로심부 지지통의 수명판정조건을 제시하기 위하여 계산한 일부이다. 수명판정을 하 기 위해서는 barrel의 강제진동 응답으로부터 fluctuating stress를 구해야만 한다. 본 연구에서는 modal analysis를 이용하여 변위를 모드함수의 급수전 개의 형태로 표시하고 가진주파수가 barrel의 고유진동수와 일치하는 모드만 을 택하여 fluctuation stress를 구하였다.

• Nuclear Engineering and Technology
• /
• 제56권2호
• /
• pp.575-600
• /
• 2024

During an earthquake on December 29th 2020, the Krško NPP automatically shutdown due to the trigger of the negative neutron flux rate signal on the power range nuclear instrumentation. From the time course of the detector signal, it can be concluded that the fluctuation in the detector signal may have been caused by the mechanical movement of the ex-core neutron detectors or the pressure vessel components rather than the actual change in reactor power. The objective of the analysis was to evaluate the sensitivity of the neutron flux at the ex-core detector position, if the detector is moved in the radial or axial direction. In addition, the effect of the core barrel movement and core inside the baffle movement in the radial direction were analysed. The analysis is complemented by the calculation of the thermal and total neutron flux gradient in radial, axial and azimuthal directions. The Monte Carlo particle transport code MCNP was used to study the changes in the response of the ex-core detector for the above-mentioned scenarios. Power and intermediate-range detectors were analysed separately, because they are designed differently, positioned at different locations, and have different response characteristics. It was found that the movement of the power range ex-core detector has a negligible effect on the value of the thermal neutron flux in the active part of the detector. However, the radial movement of the intermediate-range detector by 5 cm results in 7%-8% change in the thermal neutron flux in the active part of the intermediate-range detector. The analysis continued with an evaluation of the effects of moving the entire core barrel on the ex-core detector response. It was estimated that the 2 mm core barrel radial oscillation results in ~4% deviation in the power and intermediate-range detector signal. The movement of the reactor core inside baffle can contribute ~6% deviation in the ex-core neutron detector signal. The analysis showed that the mechanical movement of ex-core neutron detectors cannot explain the fluctuations in the ex-core detector signal. However, combined core barrel and reactor core inside baffle oscillations could be a probable reason for the observed fluctuations in the ex-core detector signal during an earthquake.

• Design & Manufacturing
• /
• 제2권3호
• /
• pp.10-15
• /
• 2008

Recently, mobile-phone with camera module has an absolute majority in the released mobile-phones. For this trend, High precision of the plastic subminiature barrel which is the core part is needed significantly because the camera module of mobile-phone must have high performance. Therefore, Structural stability of the injection mold for plastic subminiature barrel has to be guaranteed. In this paper, structural analysis of injection mold for plastic subminiature barrel is performed. Finally, the deformation trend and stability of injection mold core are analyzed.

• 한국정밀공학회:학술대회논문집
• /
• 한국정밀공학회 2002년도 추계학술대회 논문집
• /
• pp.859-862
• /
• 2002

The Core Support Barrel (CSB) is a major component of Reactor Internals, and is designed to support and protect the Reactor Core. In this study, Reactor Core, Core Shroud and CSB were simplified to coaxial cylinders and then the offset of Reactor Core & Core Shroud to the dynamic characteristic of CSB was analyzed. For the beam modes, natural frequencies of the cantilevered cylinder are compared with those of the cantilevered beam. And it was found out that shear modulus must be used correctly to convert the shell model to the equivalent beam model. From the dynamic characteristics of the beam model, it was found out that natural frequencies are proportional to the length of Reactor Core & Core Shroud and inversely proportional to the mass. From the comparison with the dynamic characteristics of a beam model and a lumped-mass model it was found out that the size of lumped-mass must be determined considering both the length and the mass of Reactor Core & Core Shroud.

• 대한기계학회논문집 C: 기술과 교육
• /
• 제5권2호
• /
• pp.127-134
• /
• 2017

원자로내부구조물은 다양한 진동하중조건에서 안전성과 건전성을 유지해야한다. 그러므로 미국원자력규제 위원회는 펌프 압력 맥동에 의한 진동을 포함하여 음향으로 유발되는 진동을 평가하기 위한 규제지침서 1.20을 제시하고 있다. 본 논문은 음향-구조 연성해석 기법을 사용하여 펌프 맥동 가진으로 인한 해석을 위해 노심지지배럴 주변의 유체와 구조의 연성을 고려하여 해석하는 방법론을 제안하였다. 해석결과는 미국 Palo Verde 1호기 종합진동평가 프로그램 발전소 시험결과와 잘 일치한다. 따라서 제안된 해석 방법론은 펌프 압력맥동에 대한 노심지지배럴의 구조응답을 평가하기 위한 효과적 방법으로 판단된다.

• 한국전산구조공학회논문집
• /
• 제15권1호
• /
• pp.59-68
• /
• 2002

원자로 내부구조물의 상부안내구조물집합체는 노심지지배럴과 내부배럴집합체와 함께 원통형의 실린더 구조이며, 유체의 난류하중과 펌프의 맥동하중으로 인한 유체유발하중을 수평방향으로 받는다. 본 논문에서는 이 유체유발하중에 대한 랜덤진동해석과 조화응답해석을 수행한 내용을 기술하였다. 이 해석을 위해 집중질량 보 요소 모델을 사용하였고, 랜덤하중과 펌프맥동하중으로 발생되는 동적응답특성을 평가하였다. 특히 원통형태의 상부안내구조물, 노심지지배럴, 내부배럴집합체 사이에서 형성되는 환형공간의 동수력 연성 효과를 고려하여 모델링 하였고, 상부안내구조물 안쪽에 설치되는 내부배럴집합체의 추가 영향을 검토하였다. 내부배럴집합체의 추가로 인한 하중조건별 최대동적응답은 구조물의 고유진동수에 영향을 받으며, 따라서 구조물의 최대동적응답은 여러 하중 조건별 동적해석 평가를 통해 보수적으로 구하여져야 한다.

• Nuclear Engineering and Technology
• /
• 제47권2호
• /
• pp.165-175
• /
• 2015

The application of neutron noise analysis (NNA) to the ex-core neutron detector signal for monitoring the vibration characteristics of a reactor core support barrel (CSB) was investigated. Ex-core flux data were generated by using a nonanalog Monte Carlo neutron transport method in a simulated CSB model where the implicit capture and Russian roulette technique were utilized. First and third order beam and shell modes of CSB vibration were modeled based on parallel processing simulation. A NNA module was developed to analyze the ex-core flux data based on its time variation, normalized power spectral density, normalized cross-power spectral density, coherence, and phase differences. The data were then analyzed with a fuzzy logic module to determine the vibration characteristics. The ex-core neutron signal fluctuation was directly proportional to the CSB's vibration observed at 8Hz and15Hzin the beam mode vibration, and at 8Hz in the shell mode vibration. The coherence result between flux pairs was unity at the vibration peak frequencies. A distinct pattern of phase differences was observed for each of the vibration models. The developed fuzzy logic module demonstrated successful recognition of the vibration frequencies, modes, orders, directions, and phase differences within 0.4 ms for the beam and shell mode vibrations.

• 한국소음진동공학회:학술대회논문집
• /
• 한국소음진동공학회 2000년도 춘계학술대회논문집
• /
• pp.679-686
• /
• 2000

Reactor Noise is defined as the fluctuations of measured instrumentation signals during full-power operation of reactor which have informations on reactor system dynamics such as neutron kinetics, thermal-hydraulics, and structural dynamics. Reactor noise analyses of ex-core neutron detector signals have been performed to monitor the vibration modes of reactor internals such as fuel assembly and Core Support Barrel in Yonggwang 3&4 Nuclear Power Plant. A real time mode separation technique have been developed and applied for the analyses. It has been found that the first vibration mode frequency of the fuel assembly was around 2.5 Hz, the beam and shell mode frequencies of CSB(Core Support Barrel) 8 Hz and 14.5 Hz, respectively. Also the analyses data base have been constructed for the continuous monitoring and diagnose of the reactor internals.

• International Journal of Fluid Machinery and Systems
• /
• 제9권4호
• /
• pp.325-331
• /
• 2016

A bathtub vortex above the outlet of a rotating barrel is simulated. By analyzing the Ekman layer theory, it can be found that the main flow circulation is inversely proportional to the thickness of Ekman layer. The thicker the Ekman boundary layer, the weaker the rotational strength and the shorter of the length of gas core is. According to this law, models of barriers with rods of different heights are established. The reduction of air-core length in this air entrainment vortex and weakening the strength of rotation field were achieved.