• 원자력산업
• /
• 제15권2호통권144호
• /
• pp.42-45
• /
• 1995

모든 원자력관련시설들은 지진으로부터 공공에 아무런 피해를 주지 않는 안전한 설계를 위하여, 일반구조물인 건물, 교량 등의 설계$\cdot$건설을 위하여 정해져 있는 내진설계법규보다도 훨씬 보수적으로 내진설계를 규정하는 원자력법규에 따라 설계 건설되고 있다. 국내의 원자력법규에 정해진 설계기준지진은 안전정지지진(Safe Shutdown Earthquake, SSE)이라 하며, 국내의 원자력발전소 및 원자력관련시설은 안정정지지진의 지반가속도 크기를 0.2g(g는 중력가속도, 1g= 1,000gal=980cm/$sec^2$)로 정하여 사용되고 있다.

• 전산구조공학
• /
• 제4권1호
• /
• pp.17-19
• /
• 1991

본 글에서는 ANS(American Nuclear Society)에서 규정하고 있는 Safety Class 구조물 중 격납건물의 내진해석모델에 대해 주로 언급하며, 기타 건물(Non-Nuclear Safety "NNS")은 일반건물과 차이점이 없음으로 언급하지 않는다. U.S. NRC(Nuclear Regulatory Commision)는 참고문헌[1]에 의해 원자력 발전소내 여러건물중 SSE(Safe Shutdown Earthquake)하중에도 견디도록 규정하여 정확한 동력학적 계산이 요구되어지나, 그 이외의 건물들은 대개 등가정적 방법에 의해 해석되어진다. 이 해석방법은 여러 빌딩규정들에서 제안된 것으로 지진에 의한 동적하중을 밑면적 단력 또는 각층에 작용하는 층전단력이라는 등가의 정적하중으로 바꿔 계산되며 이 때 사용되는 해석모델은 건물의 기본진동만을 표시할 수 있는 간단한 것이다.

• 한국압력기기공학회 논문집
• /
• 제14권2호
• /
• pp.11-15
• /
• 2018

A development plan for seismic capacity verification procedures of nuclear components based on the elastic-plastic strain (EPS) features is explained in this paper. The EPS methodology is more realistic to assess seismic responses of components to extreme seismic events beyond the safe shutdown earthquake (SSE) than current practices with the criteria of stress limits. The EPS based approach to analyze the seismic capacity of components can reduce over-conservatism in the current stress-based criteria and can incorporate the seismic responses of components deformed in plastic behavior by the motion of extreme earthquake.

• Nuclear Engineering and Technology
• /
• 제24권2호
• /
• pp.111-120
• /
• 1992

본 논문은 안전정지지진에 대한 spacer grid충격하중의 동적특성을 고찰한 것이다. 안전정지지 진 (SSE： Safe Shutodown Earthquake)을 가진력으로 사용하였고 모델로는 가장 긴 15 row의 핵연료집합체 모델을 사용하였다. 외부에서 제공되는 입력데이타의 불확실성을 보상하기 위하여 흔히 사용되는 10%증가된 입력값과, NRC 권고사항인 입력응답스펙트럼의 이동효과의 영향이 평가되었다. 평가대상으로는 spacer grid의 충격력이 사용되었다. 해석결과로서 입력응답 스펙트럼의 이동효과의 영향은 거의 무시할 수 있으나 증가된 입력값의 영향은 spacer grid의 충격력에 큰 영향을 줌을 알 수 있었다. 따라서 원자로노심의 동적응답을 위해서는 입력응답 스펙트럼을 이동시키면서 해석할 필요가 없음을 확인하였다.

• Nuclear Engineering and Technology
• /
• 제53권6호
• /
• pp.2038-2045
• /
• 2021

The main objective of this work is to propose a reliable routine standard operation procedures (SOP) for structural health monitoring and diagnosis of nuclear power plants (NPPs). At present, NPPs have monitoring systems that can be used to obtain the quantitative health record of containment (CTMT) buildings through system identification technology. However, because the measurement signals are often interfered with by noise, the identification results may introduce erroneous conclusions if the measured data is directly adopted. Therefore, this paper recommends the SOP for signal screening and the required identification procedures to identify the dynamic characteristics of the CTMT of NPPs. In the SOP, three recommend methods are proposed including the Recursive Least Squares (RLS), the Observer Kalman Filter Identification/Eigensystem Realization Algorithm (OKID/ERA), and the Frequency Response Function (FRF). The identification results can be verified by comparing the results of different methods. Finally, a preliminary CTMT healthy record can be established based on the limited number of earthquake records. It can be served as the quantitative reference to expedite the restart procedure. If the fundamental frequency of the CTMT drops significantly after the Operating Basis Earthquake and Safe Shutdown Earthquake (OBE/SSE), it means that the restart actions suggested by the regulatory guide should be taken in place immediately.

• Journal of Advanced Marine Engineering and Technology
• /
• 제27권6호
• /
• pp.790-796
• /
• 2003

As a fluid machinery for piping liquid in the reactor cooling system, multi-stage centrifugal pump requires the structural dynamic stability against external dynamic excitation. This paper is concerned with the finite element analysis of its eigen behavior and seismic response to RRS(Required Response Spectrum) curves in the case of SSE (Safe Shutdown Earthquake). Through the finite element analysis, the major vibration characteristics of multi-stage centrifugal pump(MSCP) are investigated and seismic qualification based on the IEEE codes is executed. The numerical results show that the MSCP used in this study has enough seismic strength.

• 한국소음진동공학회:학술대회논문집
• /
• 한국소음진동공학회 2005년도 추계학술대회논문집
• /
• pp.525-525
• /
• 2005

This study is for the seismic analysis and the structural integrity evaluation of the box beam for supporting nuclear fuel-transfer-basket of the HANARO reactor pool. For performing the seismic analysis and evaluating the structural integrity in air or submerged condition, the finite element model of the fuel-transfer-basket and its supporting box beam(the coupled model) was developed. The hydrodynamic effect is also considered by using added mass concept. The seismic response spectrum analyses of the coupled model under the design floor response spectrum loads of Safe Shutdown Earthquake(SSE) were performed. Through the numerical experiments, the analysis results show that the stress values of the coupled model lot the structural integrity are within the ASME Code limits.

• 한국기계가공학회지
• /
• 제13권3호
• /
• pp.42-48
• /
• 2014

The auxiliary-feed-water pump (AFWP) used to supply water during a station black out situation at nuclear power plants should meet the seismic qualification regulations stipulated in IEEE Std 323 and 344, so as to withstand earthquakes or dangerous situations. Here, we establish a model for the estimation of the structural integrity of this type of pump. If the natural frequency that results from a modal analysis is less than 33 Hz, we adopt a dynamic analysis, instead of a static analysis. A dynamic analysis was carried out taking into consideration seismic conditions such as the floor response spectra (FRS), an operation-base earthquake (OBE), and a safe-shutdown earthquake (SSE). Finally, an analytical estimation of the structural integrity of an AFWP is made through a comparison of calculated values and allowable values. If the result is less than the allowable stress, the pump is deemed to have good structural integrity. In addition, future studies will involve a stability check for rotor accidents that may occur during the operation of the pump.

• 한국기계가공학회지
• /
• 제19권4호
• /
• pp.45-51
• /
• 2020

In this study, a response spectrum analysis of an electric distributing board (EDB) was conducted to investigate seismic integrity in the design stage. For the seismic analysis, the required response spectra of a safe shutdown earthquake with 2% damping (RRS/SSE-2%) specified in GR-63-CORE Zone 4 was used as the ground spectral acceleration input. A finite element method modal analysis of the EDB was also performed to examine the occurrence of resonance within the frequency range of the earthquake response spectrum. Furthermore, static stress caused by deadweight was analyzed. The resultant total maximum stress of the EDB structure was calculated by adding the maximum stresses from both seismic and static loads using the square root of the sum of the squares (SRSS) method. Finally, the structural safety of the EDB was investigated by comparing the resultant total maximum stress with the allowable stress.

• 한국안전학회지
• /
• 제31권6호
• /
• pp.129-134
• /
• 2016

Standard design of APR+(advanced power reactor plus) was certified at 2014 by Korea regulatory body. Based on the experience gained from OPR1000 and APR1400, the APR1400 was being developed as a 1,500MWe class reactor using Korean technologies for design code, reactor coolant pump, and man-machine interface system. APR+ has been basically designed to have the seismic design basis of safe shutdown earthquake (SSE) 0.3g, a 4-train safety concept based on N+2 design philosophy, and a passive auxiliary feedwater system (PAFS). Also, safety issues on the Fukushima-type accidents have been extensively reviewed and applied to enhance APR+ safety. APR+ provides higher reliability and safety against tsunami and earthquake. The purpose of this paper is to implement probabilistic safety assessment considering these design features and to analyze sensitivity of core damage frequency for large loss of coolant accident of APR+.