• Nuclear Engineering and Technology
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• v.27 no.5
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• pp.710-720
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• 1995

The propagation of pump-induced pressure pulsation in a reactor is important because of the potential for vibration and resultant damage of reactor internals. A hydrodynamic model has been developed to obtain the pressure fluctuation due to the operation of pumps in the annulus(between the core support barrel and reactor vessel of a pressurized water reactor) including the coolant inlet pipe. The mathematical analysis is formulated in accordance with the linearized Navier-Stokes equation by assuming a compressible, inviscid flow. Two regions are considered separately and by coupling the solutions of the inlet pipe and the annulus, the inlet nozzle pressure(pressure at pipe and annulus interface) is to be calculated without assumptions. The geometric parameter effect on the pump-induced pressure pulsation is evaluated. Comparison of predicted and measured inlet nozzle pressure values for each forcing frequency shows good order of magnitude agreement.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.21 no.8
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• pp.1270-1275
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• 1997

This paper concerns with the finite element analysis of reactor internals with structural faults. For investigating the influence of hold-down spring faults on dynamic characteristics of CSB (core support barrel), reactor internals of Ulchin-1 nuclear power plant are modeled using finite element method and simulated with artificial defects on the hold-down springs. To prove the validity of the finite element models, the calculated natural frequencies of CSB in normal state are compared with those from the measurement results, which shows good agreement. According to results of finite element analysis, CSB beam mode natural frequency decreases by 4.5% in the case of 10% partial relaxation of hold-down springs, and decreases by 18.4% in the case of 20%. The range of shell mode natural frequency change is within 5.3%.

• Computational Structural Engineering
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• v.8 no.1
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• pp.13-18
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• 1995

원전의 주요 구조물은 극심한 재해를 초래할 수 있는 자연현상(지진, 쓰나미 등)이 수반하는 극한하중에 견딜 수 있어야 하고 내부의 안전관련 기기의 기능과 만일의 사고(주증기나 고에너지 배관파단사고)시에도 구조적 안전성이 보장되도록 설계.시공되고 있다. 또한 운전원의 조작실수나 오판을 극복할 수 있도록 원자로 고유의 안전성과 자연법칙에 의한 정상유지를 할 수 있는 신기술이 도입되고 있다. 품질보증 계획에 의한 최선의 품질보장, 엄격한 방사선 감시와 온배수 영향의 최소화, 최신방사성 폐기물 처분기술 적용 등을 통하여 기술의존형 에너지인 원자력의 활용증대가 불가피하다고 판단된다.

• Journal of Energy Engineering
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• v.24 no.3
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• pp.135-141
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• 2015

The core melts stratifies into lower mixture layer and upper metal layer by density in a severe accident condition. The decay heat generated from the mixture layer threatens the integrity of the reactor vessel. This study simulated the natural convection heat transfer of the mixture layer with volumetric heat source using the mass transfer system. $H_2SO_4-CuSO_4$ electroplating system was used as the mass transfer system. With the modified Rayleigh number of $3{\times}10^{14}$, the Nusselt number showed minimum at the bottom and increased along curvature to the top of the experimental apparatus.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.15 no.2
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• pp.58-63
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• 2019

This paper presents preliminary analysis and results on IASCC sensitivity of a core shroud in the reactor pressure vessel. First, neutron irradiation flux distribution of the reactor internals was calculated by using the Monte Carlo simulation code, MCNP6.1 and the nuclear data library, ENDF/B-VII.1. Second, based on the neutron irradiation flux distribution, temperature and stress distributions of the core shroud during normal operation were determined by performing finite element analysis using the commercial finite element analysis program, ABAQUS, considering irradiation aging-related degradation mechanisms. Last, IASCC sensitivity of the core shroud was assessed by using the IASCC sensitivity definition of EPRI MRP-211 and the finite element analysis results. As a result of the preliminary analysis, it was found that the point at which the maximum IASCC sensitivity is derived varies over operating time, initially moving from the shroud plate located in the center of the core to the top shroud plate-ring connection brace over operating time. In addition, it was concluded that IASCC will not occur on the core shroud even after 60 years of operation (40EFPYs) because the maximum IASCC sensitivity is less than 0.5.

• Journal of the Computational Structural Engineering Institute of Korea
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• v.15 no.1
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• pp.59-68
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• 2002

The cylindrical Upper Guide Structure assembly of the reactor intervals wish the Core Support Barrel and the Inner Barrel Assembly is subjected to flow induced loads horizontally which include random pressure fluctuation due to turbulent flow and pump pulsation pressures. The purpose of this papers is to perform random vibration and harmonic response analyses fort flow induced loads. The dynamic response characteristics due to random turbulence and pump pulsation loads were evaluated using the lumped mass beam model. Especially the model considered the annulus effects due to water gaps existing between cylindrical structures such as the Upper Guide Structure Barrel, the Core Support Barrel, and the Inner Barrel Assembly. The effect of the Inner Barrel Assembly inside the Upper Guide Structure assembly was studied. The peak dynamic responses lot each loading condition due to the addition of IBA were affected by the natural frequencies of the structures. Therefore the peak dynamic responses of the structures should be conservatively obtained from evaluation of dynamic analysis for various loading conditions.

• Journal of the KSME
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• v.19 no.3
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• pp.190-196
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• 1979

본 강좌에서는 강도와 밀접한 관계를 가지고 있는 파양문제에 대하여, 특히 파괴된 파단면을 중심으로 기초적인 일반사항을 기술해 보고져한다. 그 내용으로는 파괴와 Fractograph, 전자현 미경과 Peplica 법, Shadowing의 유효성, 파괴의 종류와 Fractograph의 일열 및 파단면형성에 관한 기구등이다. 물체가 파괴된다고 하는 것은 외력을 가하면 물체가 2개로 나누어진다는 단 순한 현상이라고 생각되기 쉽지만, 실제로는 물체가 왜, 이렇게 파괴되는가에 대해서는 알 수 없는 것이 대단히 많다. 고체가 2개로 분리되는 macro 현상은 원자배열이나 응집력등과 같은 micro적인 것에 기인하는 복잡한 것들이 있다. 이미 여러분들이 아는 바와같이, 구조물이나 기 계의 파괴는 돌이킬 수 없는 사고에 이르기 쉽다. 작은 부품 1개의 파괴라던지, 강판의 흠이 원 인이 되어 항공기나 선박등의 사고가 생기는 경우가 있다. 금후에는 원자로나 핵융합반응장치 등이 인류의 에너지원으로서 많이 이용될 가능성이 있지만, 파괴사고가 허용되어서는 안된다. 과학기술의 진보와 더불어, 기계 및 구조물에는 보다 가한 재료가 요구되고, 개발되어가고 있다. 따라서 충분히 안전하게 설계되어 있다고 생각할 수 있는 구조라 하더라도 재료내부의 결함을 기점으로 하여 가끔 파괴가 일어남은 부인하지 못할 사실이다.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.19 no.1
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• pp.1-5
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• 2023

This paper describes a newly-developed analysis methodology in comprehensive vibration assessment program (CVAP) of reactor internals to develop a valid-prototype for the design of nuclear power plants. The new analysis methodology developed in this study will be confirmed through a scale model testing (SMT). Based on the measurements obtained from dynamic pressure transducers in the SMT, a new non-dimensional equation is developed to apply the forcing functions at reactor internals for the prototype. In addition to the new non-dimensional equation, a computational fluid dynamics(CFD) is used to develop the application of the hydraulic loads at reactor internals for the prototype.

• Journal of the Earthquake Engineering Society of Korea
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• v.5 no.5
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• pp.25-33
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• 2001

For the seismic design and analysis of LMR(liquid metal reactor) being developed in Korea, it is necessary to develop the simple seismic analysis model including the fluid-structure interaction effects. In this paper, the theoretical backgrounds for the fluid added mass of the immersed concentric cylinders are investigated and the seismic analysis code using the Runge-Kutta algorithm, which can consider the fluid added mass matrix in system matrix, are developed to perform the time history seismic analysis. Form the coupled modal analysis and the seismic analysis for the simple immersed concentric cylinders, it is verified that the fluid added mass significantly affect the vibration characteristics and the seismic responses. Therefore the fluid coupled effects should be carefully considered in seismic response analysis of the immersed concentric cylinders.

• Computational Structural Engineering
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• v.5 no.3
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• pp.105-112
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• 1992

The design of reactor internals requires the accurate vibration characteristics of each component for subsequent dynamic structural response analysis. For Korean standard nuclear power plant some modifications on the Control Element Assembly shroud from the reference design have been made. Since the shroud is complex in geometry having an array of vertical round tubes and webs in a square grid pattern, and being tied down by preloaded tie rods into position, it is planned to perform a vibration measurement program consisting of both experimental and analytical modal studies upon that component. To determine the proper test conditions, the pre-test analysis has been performed using the general purpose structural analysis program ANSYS. Also the effects of the number of master degrees of freedom, holes in the web and tie-rod preload on the natural frequencies are examined prior to the pre-test analysis. After decision of appropriate finite element model, frequency analysis and harmonic analysis are performed and ideas for the test conditions such as the number of measurement points, their locations, measurement frequency range and the excitation force level are determined.