• Nuclear Engineering and Technology
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• 제47권6호
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• pp.766-775
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• 2015

The integrity assessment of reactor vessel internals should be conducted in the design process to secure the safety of nuclear power plants. Various loads such as self-weight, seismic load, flow-induced load, and preload are applied to the internals. Therefore, the American Society of Mechanical Engineers (ASME) Code, Section III, defines the stress limit for reactor vessel internals. The present study focused on structural response analyses of the upper guide structure upper flange. The distributions of the stress intensity in the flange body were analyzed under various design load cases during normal operation. The allowable stress intensities along the expected sections of stress concentration were derived from the results of the finite element analysis for evaluating the structural integrity of the flange design. Furthermore, seismic analyses of the upper flange were performed to identify dynamic behavior with respect to the seismic and impact input. The mode superposition and full transient methods were used to perform time-history analyses, and the displacement at the lower end of the flange was obtained. The effect of the damping ratio on the response of the flange was also evaluated, and the acceleration was obtained. The results of elastic and seismic analyses in this study will be used as basic information to judge whether a flange design meets the acceptance criteria.

• 한국지진공학회:학술대회논문집
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• 한국지진공학회 1998년도 춘계 학술발표회 논문집 Proceedings of EESK Conference-Spring 1998
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• pp.147-154
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• 1998

The purpose of this paper is to propose the draft guidelines of seismic isolation design of Liquid Metal Reactor (LMR) using high damping laminated rubber bearings. The scopes of guidelines include design requirements of a seismically isolated system and components, seismic isolator, isolation system, interface system between seismic isolation and non-seismic isolation part, qualification and acceptance tests of seismic isolator, seismic isolation reliability, and seismic safety and monitoring system. Proposed guidelines shall be revised to extend to general design guideline for nuclear facilities by further research and discussions.

• Nuclear Engineering and Technology
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• 제54권12호
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• pp.4571-4584
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• 2022

The analysis of rapid flow transients in Reactor Coolant Pumps (RCP) is essential for a reactor safety study. An accurate and precise analysis of the RCP coastdown is necessary for the reactor design. The coastdown of RCP affects the coolant temperature and the colloidal crud in the primary coolant. A realistic and kinetic model has been used to investigate the behavior of activated colloidal crud in the primary coolant and steam generator that solves the pump speed analytically. The analytic solution of the non-dimensional flow rate has been determined by the energy ratio β. The kinetic energy of the coolant fluid and the kinetic energy stored in the rotating parts of a pump are two essential parameters in the form of β. Under normal operation, the pump's speed and moment of inertia are constant. However, in a coastdown situation, kinetic damping in the interval has been implemented. A dynamic model ACCP-SMART has been developed for System Integrated Modular and Advanced Reactor (SMART) to investigate the corrosion due to activated colloidal crud. The Fickian diffusion model has been implemented as the reference corrosion model for the constituent component of the primary loop of the SMART reactor. The activated colloidal crud activity in the primary coolant and steam generator of the SMART reactor has been studied for different equilibrium corrosion rates, linear increase in corrosion rate, and dynamic RCP coastdown situation energy ratio b. The coolant specific activity of SMART reactor equilibrium corrosion (4.0 mg s^-1) has been found 9.63×10^-3 µCi cm^-3, 3.53×10^-3 µC cm^-3, 2.39×10^-2 µC cm^-3, 8.10×10^-3 µC cm^-3, 6.77× 10^-3 µC cm^-3, 4.95×10^-4 µC cm^-3, 1.19×10^-3 µC cm^-3, and 7.87×10^-4 µC cm^-3 for ²⁴Na, ⁵⁴Mn, ⁵⁶Mn, ⁵⁹Fe, ⁵⁸Co, ⁶⁰Co, ⁹⁹Mo, and ⁵¹Cr which are 14.95%, 5.48%, 37.08%, 12.57%, 10.51%, 0.77%, 18.50%, and 0.12% respectively. For linear and exponential coastdown with a constant corrosion rate, the total coolant and steam generator activity approaches a higher saturation value than the normal values. The coolant and steam generator activity changes considerably with kinetic corrosion rate, equilibrium corrosion, growth of corrosion rate (ΔC/Δt), and RCP coastdown situations. The effect of the RCP coastdown on the specific activity of the steam generators is smeared by linearly rising corrosion rates, equilibrium corrosion, and rapid coasting down of the RCP. However, the time taken to reach the saturation activity is also influenced by the slope of corrosion rate, coastdown situation, equilibrium corrosion rate, and energy ratio β.

• 한국전산구조공학회논문집
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• 제26권2호
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• pp.131-138
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• 2013

면진장치는 상부구조물의 지진력을 감소시키는데 크게 기여하지만, 고감쇠고무 적층받침에 사용되는 고무재료는 시간이 경과함에 따라 열화되어 상부구조물의 동특성과 기기들의 지진응답에 영향을 줄 수 있다. 따라서 면진장치의 경년열화를 고려한 구조물의 지진응답을 분석하는 연구가 필요하다. 본 연구에서는 기존 문헌을 통하여 분석된 고무의 경년열화 특성을 사용하여 면진장치를 모델링하였다. 면진된 원전의 지진응답을 평가하기 위하여 격납건물과 보조건물을 대상 구조물로 선정하고, 진동수 성분이 다양한 입력지진동을 사용하여 구조물의 고유진동수, 최대지진응답, 층응답스펙트럼을 시간의 경과에 따라 분석하였다. 해석결과에 의하면 면진장치의 경년열화에 의하여 지진응답이 소폭 증가하였으며, 면진장치가 설치된 후 20년까지 지진응답의 증가율이 크게 나타나므로 이 기간에 상세한 검사가 시행되어야 할 것이다.

• 한국지진공학회논문집
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• 제7권2호
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• pp.21-27
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• 2003

일반적으로 유체-구조물 상호작용을 고려한 유체속 구조물들의 지진 및 진동해석에는 주어진 시스템에 대한 유체부가질량을 추정하여 구조물관 연계하는 단순해석 방법을 주로 사용한다. 실제로 유체속 구조물의 응답특성은 유체부가질량 뿐만 아니라 유체점성으로 인한 감쇠영향을 받으며 이들은 모두 연계항을 갖는 복잡한 행렬 형태로 나타난다. 본 연구에서는 비점성 및 점성 유체에 대한 Navier-Stokes 지배방정식의 선형화를 통한 유한요소 정식화를 유도하였다. 이를 이용하여 유한요소 해석 프로그램을 작성하고 6각형 단면특성을 갖는 액체금속로 노심에 대하여 덕트집합체 사이의 유체간격과 레이놀즈수 변화에 따른 유체부가질량과 유체감쇠에 대한 유한 요소 해석을 수행한 결과, 유체간격이 줄어들수록 유체부가질량은 유체점성의 영향을 크게 받고 유체감쇠는 점성으로 인하여 레이놀즈수의 영향을 크게 받는 것으로 나타났다. 또한 편심을 갖는 동축원통에 대한 유한요소 해석결과, 편심이 증가할수록 유체부가질량은 크게 증가하지만 유체감쇠는 편심이 작은 경우 거의 변화가 없으며 어느 일정 수준이상으로 편심이 커질 경우에는 크게 영향을 받는 것으로 나타났다.

• 한국소음진동공학회:학술대회논문집
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• 한국소음진동공학회 2003년도 춘계학술대회논문집
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• pp.82-87
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• 2003

Nuclear fuel with a big slenderness ratio is susceptible to flow-induced vibration under very severe conditions of high temperature, high flow and exposure to irradiation in nuclear reactor. The fuel assembly should, therefore, be designed to escape any resonance due to the vibration during the reactor operation, in particular, in case of the design changes. In addition, the amplitudes due to the grid vibration, the fuel rod vibration and the fuel assembly vibration should be minimized to reduce the grid-to-rod fretting wear. Fuel assembly vibration tests in air at room temperature and in water at high temperature have been performed to investigate fuel vibration behaviors. The frequency and damping during the test in air have been compared to those in water. Through the hydraulic test, the advanced assembly has been evaluated not to be susceptible to any resonance. In addition, the test data from the tests can be used to make fuel model and to evaluate grid-to-rod fretting wear.

• 대한전기학회논문지
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• 제38권11호
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• pp.861-869
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• 1989

본 논문에서는 폰트라이긴의 최대원리를 이용한 가압경수형 원자로(PWR)의 부하추종 운전시 제논진동 최적화 문제가 제시되었다. 최적화 모델은 2차 목적함수를 갖고 있는 최적 추적제어문제로 정식화 하였으며, 1군 확산방정식과 제논-아이오다인 동특성 방정식을 등호 제약조건으로 고려하였다. 최적화 모델에 최대원리를 적용하므로서, 문제는 제약조건이 없는 단일시간 문제로 분리되었으며, 분리된 부 문제는 공액 경사법을 이용하여 최적화 하였다. 계산결과는 제논진동이 최소화되어 원자로가 규정된 출력분포를 유지하면서 전력계총에서 요구하는 출력을 잘 추종 하였다.

• 한국지진공학회논문집
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• 제25권2호
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• pp.83-91
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• 2021

Seismic designs for Korean nuclear power plants (NPPs) under earthquakes' design basis are noticed due to the recent earthquake events in Korea and Japan. Japan has developed the technologies and experiences of the NPPs through theoretical research and experimental verification with extensively accumulated measurement data. This paper describes the main features of the design-time history complying with the Japanese seismic design standard. Proper seed motions in the earthquake catalog are used to generate one set of design time histories. A magnitude and epicentral distance specify the amplitude envelope function configuring the shape of the earthquake. Cumulative velocity response spectral values of the design time histories are compared and checked to the target response spectra. Spectral accelerations of the time histories and the multiple-damping target response spectra are also checked to exceed. The generated design time histories are input to the reactor building seismic analyses with fixed-base boundary conditions to calculate the seismic responses. Another set of design time histories is generated to comply with Korean seismic design procedures for NPPs and used for seismic input motions to the same reactor containment building seismic analyses. The responses at the dome apex of the building are compared and analyzed. The generated design time histories will be also applied to subsequent seismic analyses of other Korean standard NPP structures.

• KIEE International Transactions on Power Engineering
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• 제11A권4호
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• pp.9-14
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• 2001

This paper reports a set of digital time-domain simulation studies conducted on 154 kV wound Potential Transformer(PT) int he 154 kV Gas Insulated Substation(GIS). The Electro-Magnetic Transient Program(EMTP) is used to develop the PT model and conduct the transient studies. The accuracy of the PT model is verified through comparison of the EMTP simulation results with those obtained from the field test results. The investigations shows that the developed model can accurately predict PT transient resonance, especially, the phenomenon of ferroresonance. The model is developed not only to determine impact of transients on PT response but also to design ferroresonance suppressor devices of PT. And it can also be used to predict PT transient response on power system monitoring and protection scheme.

• 한국지진공학회:학술대회논문집
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• 한국지진공학회 1997년도 추계 학술발표회 논문집 Proceedings of EESK Conference-Fall 1997
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• pp.117-125
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• 1997

ABSTRACT This paper presents results of parametric studies of the seismic responses of a reactor containment structure on layered base soil. Among the numerous parameters, this study concentrates on the effects of embedment of structure to the base soil, thickness of the soil layers, stiffness of the base soil, and the definition point of the input motion. For the analysis, a substructure method using frequency independent impedances is adopted. The method is based on the mode superposition method in time domain using the composite modal damping values of the SSI system computed from the ratio of dissipated energy to the strain energy for each mode. From the study results, the sensitives of each parameter on the earthquake responses have been suggested for the practical application of the substructure method of SSI analysis.