• 한국소음진동공학회논문집
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• 제21권1호
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• pp.47-55
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• 2011

Reactor vessel internals comprehensive vibration assessment program(RVI CVAP) is one of the necessary tests to ensure the safety of nuclear power plants. RVI CVAP of U.S. nuclear regulatory commission regulatory guide 1.20(U.S. NRC R.G. 1.20) consists of the analysis, measurement and inspection. One of the core technologies of the measurement program for RVI CVAP is to select suitable sensors because the measurement is conducted during the critical path of the construction period of nuclear power plants. Therefore, we analyzed RVI thermal-hydraulic and structure design data of Palo Verde nuclear power plant(U.S.), Yonggwang nuclear power plant(Korea) and APR1400 and researched measuring sensors used in them; moreover, we investigated sensors used for measurement of RVI CVAP for the last 20 years throughout the world. Based on these results, we selected suitable measuring sensors for RVI CVAP in advanced power reactor 1400(APR1400).

• Structural Engineering and Mechanics
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• 제58권1호
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• pp.39-58
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• 2016

Since the publication of ACI 318-02, the concrete capacity design (CCD) method has been used to determine the resistance of unreinforced concrete anchors. The regulation of steel-reinforced anchors was proposed in ACI 318-08. Until ACI 318-08, the shear resistance of concrete breakout for an unreinforced anchor during an earthquake was reduced to 75% of the static shear strength, but this reduction has been eliminated since ACI 318-11. In addition, the resistance of a hairpin-reinforced anchor was calculated using only the strength of the steel, and a regulation on the dynamic strength was not given for reinforced anchors. In this study, shaking table tests were performed to evaluate the dynamic shear strength of unreinforced and hairpin-reinforced cast-in-place (CIP) anchors during earthquakes. The anchors used in this study were 30 mm in diameter, with edge distances of 150 mm and embedment depths of 240 mm. The diameter of the hairpin steel was 10 mm. Shaking table tests were carried out on two specimens using the artificial earthquake, based on the United States Nuclear Regulatory Commission (US NRC)'s Regulatory Guide 1.60, and the Northridge earthquake. The experimental results were compared to the current ACI 318 and ETAG 001 design codes.

• 한국지진공학회:학술대회논문집
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• 한국지진공학회 2001년도 추계 학술발표회 논문집 Proceedings of EESK Conference-Fall 2001
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• pp.67-74
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• 2001

This paper presents the Safe Shutdown Earthquake(SSE) input motion for the seismic design of the Advanced Power Reactor 1400(APR1400). The Design Ground Response Spectra(DGRS) far the SSE is based on the design spectrum specified in regulatory Guide(RG) 1.60 of U.S. Nuclear Regulatory Commission(US NRC), anchored to a Peak Ground Acceleration(PGA) of 0.3g and enriched in the high frequency range. This SSE seismic input motion is to be applied to the seismic analysis as the free-field seismic motion at the ground surface of both the rock and generic soil sites fur APRI1400. The enrichment for APR1400 seismic input motion is performed considering the current US NRC regulations, the seismic hazard studies performed by the Lawrence Livermore National Laboratory (LINL) and Electric Power Research Institute(EPRI) for the Central and Eastern United States nuclear power plant sites, and the seismic input motions used in the design certifications of the three existing U.S. advanced standard plants. It is represented by a set of DGRS and the accompanying Target Power Spectral Density(PSD) Function in both the horizontal and vertical directions.

• 한국소음진동공학회:학술대회논문집
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• 한국소음진동공학회 2012년도 춘계학술대회 논문집
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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.

• 한국소음진동공학회논문집
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• 제22권5호
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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.

• 한국소음진동공학회:학술대회논문집
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• 한국소음진동공학회 2014년도 추계학술대회 논문집
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• pp.99-101
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• 2014

A Comprehensive Vibration Assessment Program (CVAP) for steam generator internals (SGI) of Advanced Power Reactor 1400 (APR1400) is being performed in accordance with the United States Nuclear Regulatory Commission (U.S. NRC) Regulatory Guide 1.20 (RG 1.20) revision 3. This paper studies the vibration characteristics of moisture separator assembly as part of the vibration and stress analysis program for APR1400 SGI CVAP. The natural frequencies, mode shapes, and structural behavior of moisture separator assembly were investigated through modal analysis using finite element method and experimental measurement. Since the moisture separator consists of several items with complicated shape, an idealized shell model was used in the finite element analysis. Group of local modes caused by moisture separators and significant modes of shroud and separator support plate were identified. The results of this paper are to be utilized in the structural response analysis of moisture separator assembly.

• Nuclear Engineering and Technology
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• 제45권2호
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• pp.249-256
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• 2013

In accordance with the US Nuclear Regulatory Commission (US NRC), Regulatory Guide 1.20, the reactor vessel internals comprehensive vibration assessment program (RVI CVAP) has been developed for an Advanced Power Reactor 1400 (APR1400). The purpose of the RVI CVAP is to verify the structural integrity of the reactor internals to flow-induced loads prior to commercial operation. The APR1400 RVI CVAP consists of four programs (analysis, measurement, inspection, and assessment). Thoughtful preparation is essential to the measurement program, because data acquisition must be performed only once. The optimized design of a vibration and stress measurement system for the RVI CVAP is essential to verify the integrity of the APR1400 RVI. We successfully designed a vibration and stress measurement system for the APR1400 RVI CVAP based on the design materials, the hydraulic and structural analysis results, and performance tests of transducers in an extreme environment. The measurement system designed in this paper will be utilized for the APR1400 RVI CVAP as part of the first construction project in Korea.

• Nuclear Engineering and Technology
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• 제52권5호
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• pp.995-1001
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• 2020

With the development of digital instrumentation and control (I&C) devices, cyber security at nuclear power plants (NPPs) has become a hot issue. The Stuxnet, which destroyed Iran's uranium enrichment facility in 2010, suggests that NPPs could even lead to an accident involving the release of radioactive materials cyber-attacks. However, cyber security research on industrial control systems (ICSs) and supervisory control and data acquisition (SCADA) systems is relatively inadequate compared to information technology (IT) and further it is difficult to study cyber-attack taxonomy for NPPs considering the characteristics of ICSs. The advanced research of cyber-attack taxonomy does not reflect the architectural and inherent characteristics of NPPs and lacks a systematic countermeasure strategy. Therefore, it is necessary to more systematically check the consistency of operators and regulators related to cyber security, as in regulatory guide 5.71 (RG.5.71) and regulatory standard 015 (RS.015). For this reason, this paper attempts to suggest a template for cyber-attack taxonomy based on the characteristics of NPPs and exemplifies a specific cyber-attack case in the template. In addition, this paper proposes a systematic countermeasure strategy by matching the countermeasure with critical digital assets (CDAs). The cyber-attack cases investigated using the proposed cyber-attack taxonomy can be used as data for evaluation and validation of cyber security conformance for digital devices to be applied, and as effective prevention and mitigation for cyber-attacks of NPPs.

• 한국강구조학회 논문집
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• 제22권5호
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• pp.487-496
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• 2010

미국의 NRC(United States Nuclear Regulatory Commission)에서 발간된 "Regulatory Guide 1.61 of United States NRC(2007)"는 원전구조물의 내진설계에 적용되는 구조감쇠비를 철근콘크리트(이하 RC)구조의 경우 4%(OBE)와 7%(SSE), 강구조의 경우 3%(OBE)와 4%(SSE)를 규정하고 있다. 그러나 최근 개발된 강판-콘크리트(이하 SC)구조의 경우 구조감쇠비에 대한 규정이 없다. 본 연구의 목적은 RC구조와 SC구조의 감쇠비의 상대적 차이를 비교함으로서 SC구조의 감쇠비를 조사하는 것이다. 4개의 실험체, RC-S, RC-M, SC-S 그리고 SC-M에 대한 실험적 연구가 수행되었다. 뒷 글자 S와 M은 실험체의 거동이 전단에 의해 지배되는 것과 휨에 의해 지배되는 것을 의미한다. 실험방법은 엑츄에이터와 실험체의 질량사이를 연결하는 인장시험편이 파단되면서 실험체의 자유진동을 발생하게 하는 방법을 적용하였다. 가속도계를 이용하여 측정된 실험데이타를 분석하여 하중의 크기에 따른 기본진동수와 감쇠비를 결정하였다. 4개 실험체의 감쇠비를 비교분석하여 SC구조의 감쇠비는 OBE해석에 RC구조와 동일하게 4%를 제안하였으며 SSE해석의 경우 RC구조의 감쇠비보다 1% 적은 6%를 제안하였다.

• Journal of Radiation Protection and Research
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• 제38권2호
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• pp.60-67
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• 2013

가우시안 플륨모델(Gaussian plume model)을 사용한 대기확산의 예측에서 확산계수는 결과에 중요한 영향을 미치는 변수이다. 확산계수의 평가방법은 다양하며, 본 연구에서는 미국 원자력규제위원회(U. S. NRC) 권고 규제지침, 캐나다 원자력안전위원회(CNSC) 권고 규제지침, 확률론적 사고결말해석코드 MACCS와 MACCS2에서 권고 또는 적용하는 방법을 고찰하였다. U. S. NRC에서 권고하는 부지적합성 평가를 위한 가상사고시 대기확산모델을 기반으로 확산계수의 평가방법이 대기확산인자에 미치는 영향을 분석하였다. 확산계수는 Pasquill-Gifford 곡선을 기반으로 각기 다른 연구자들에 의해 얻어진 곡선의 피팅식(curve fitting equations)을 적용 또는 권고하고 있음을 확인하였다. 수평확산계수는 모든 규제지침과 코드에서 플륨의 사행효과를 반영하여 보정하고 있으나 그 적용 방법에 있어서는 차이를 나타냈다. 수직확산계수는 U. S. NRC 권고 규제지침을 제외하고 표면거칠기를 반영하여 보정하고 있다. 특정 표면거칠기에 대해 확산계수의 적용방법에 따라 대기확산인자는 최대 약 4배의 차이를 나타냈다. 표면거칠기는 대기확산인자에 중요한 영향을 나타냈으며, 동일 적용방법에 대해 표면거칠기에 따라 대기확산인자는 약 2~3배의 차이를 나타냈다.