• 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).

• 한국안전학회지
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• 제37권5호
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• pp.74-79
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• 2022

This study analyzes the contents of the second revised Regulatory Guide 1.180 (Revision 2) for electromagnetic compatibility qualification published by the U. S. Nuclear Regulatory Commission by comparing them with those of the previous version. The methods and acceptance criteria of both CE101 and CE102 tests pertaining to conductive emission and RE102 test for radiation emission are observed to have been modified in Revision 2. Furthermore, the revised guide is found to afford flexibility in using alternative methods for electromagnetic interference qualification by allowing the combination of different technical base standards.

• Nuclear Engineering and Technology
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• 제56권1호
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• pp.309-316
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• 2024

The U.S. Nuclear Regulatory Commission (NRC) Regulatory Guide (RG) 1.20 provides guidance on the comprehensive vibration assessment program (CVAP) to be performed on reactor internals during preoperational and startup tests. The purpose of the program is to identify loads that could cause vibration in the reactor internals and to ensure that these vibrations do not affect their structural integrity. The structural vibrational analysis program involves creating finite element analysis models of the reactor internals and calculating their structural responses when subjected to vibration loads. The appropriateness of the structural analysis methodology must be demonstrated through benchmarks or any other reasonable means. Although existing structural analysis methodologies have been proven to be appropriate and are widely used, this paper presents the development of an improved new structural analysis methodology for APR1400 reactor internals using scaled model tests.

• Nuclear Engineering and Technology
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• 제50권6호
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• pp.963-970
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• 2018

The U.S. Nuclear Regulatory Commission, Regulatory Guide 1.20 needs to perform a comprehensive vibration assessment program for reactor internals during preoperational and startup testing for nuclear power plants and extended power uprate. Although the measurement program is one of the core programs, it is rarely carried out except for a first-of-a-kind or a unique design. This article describes measurement results of vibration and stress for the comprehensive vibration assessment program for an APR-1400 reactor internals. The measurement was performed at an upper guide structure during the pre-core hot functional test of Shin Kori unit 4 reactor internals because the Shin Kori unit 3 and 4 are the first construction project for the APR-1400, and the upper guide structure assembly was to design change compared with the valid prototype. We confirmed that all measured results are within the test acceptance criteria. It means that the structural integrity of the APR-1400 reactor internals was secured for the flow-induced vibration.

• 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.

• Nuclear Engineering and Technology
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• 제49권3호
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• pp.651-661
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• 2017

The Korea Institute of Nuclear Safety, as a representative organization of Korea, in February 2012 participated in an international Program to Assess the Reliability of Emerging Nondestructive Techniques initiated by the U.S. Nuclear Regulatory Commission. The goal of the Program to Assess the Reliability of Emerging Nondestructive Techniques is to investigate the performance of emerging and prospective novel nondestructive techniques to find flaws in nickel-alloy welds and base materials. In this article, Korean round-robin test results were evaluated with respect to the test blocks and various nondestructive examination techniques. The test blocks were prepared to simulate large-bore dissimilar metal welds, small-bore dissimilar metal welds, and bottom-mounted instrumentation penetration welds in nuclear power plants. Also, lessons learned from the Korean round-robin test were summarized and discussed.

• 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.

• 한국소음진동공학회논문집
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• 제25권9호
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• pp.599-605
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• 2015

U.S. Nuclear Regulatory Commission(NRC), Regulatory Guide(RG) 1.20(Rev.3, 2007년)은 원자로 냉각재계통, 주증기, 주급수 및 복수시스템의 주요 배관 및 기기에 대하여 압력변동 및 진동에 의한 잠재적 유해효과에 대한 평가를 요구한다. 그러나 증기발생기와 연결된 주증기, 주급수 및 복수시스템의 주요 배관 전체에 대하여 상세 해석하는 것은 매우 복잡하여 한계가 있다. 이 논문은 APR1400 원전의 종합진동평가(comprehensive vibration assessment program, CVAP)를 수행하기 위하여 증기발생기에 연결된 2차측 주요 배관의 음향공진과 펌프유발진동을 위한 간이평가 방법에 관한 것이다. 이 논문에서는 이러한 배관시스템의 잠재적 진동 원인이 무엇인지, 음향공진과 펌프유발진동의 가능성을 예방하기 위한 간이평가 방법은 무엇인지를 고찰하고자 한다. 이 논문은 APR1400 원전 증기발생기와 연결된 주증기 및 주급수 배관의 유동유발진동 간이평가를 위해 사용될 것으로 판단된다.

• Nuclear Engineering and Technology
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• 제50권5호
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• pp.709-716
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• 2018

A criticality safety analysis was performed for the APR-1400 spent fuel pool region-II to ensure the safe storage of spent fuel, with credit taken for depletion and in-rack neutron absorbers (Metamic panels). PLUS7 fuel assembly was modeled using TRITON-NEWT of SCALE-6.1. The burnup-dependent cross-section library was generated under limiting core-operating conditions with 5%-w U-235 initial enrichment. MCNP5 was used to evaluate the neutron multiplication factor in an infinite array of rack cells with the axially nonuniformly burnt PLUS7 assemblies under normal, abnormal, and accident conditions; including all biases and uncertainties. The main purpose of this study is to investigate reactivity variations due to the critical depletion and reactor operation parameters. The approach, assumptions, and modeling methods were verified by analyzing the contents of the most important fissile and the associated reactivity effects. The Nuclear Regulatory Commission (NRC) guidance on k-eff being less than 1.0 for spent fuel pools filled with unborated water was the main criterion used in this study. It was found that assemblies with 49.0 GWd/MTU and 5.0 w/o U-235 initial enrichment loaded in Region-II satisfy this criterion. Moreover, it was found that the end effect resulted in a positive bias, thus ensuring its consideration.

• 방사성폐기물학회지
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• 제21권3호
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• pp.347-357
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• 2023

United States Nuclear Regulatory Commission (U.S. NRC) specifies regulations on obtaining licenses and describes the technical position on the average waste concentration, also known as Concentration Averaging and Encapsulation Branch Technical Position (CA BTP); CA BTP helps classify blendable waste and discrete items and address concentration averaging. The technical position details are reviewed and compared in a real environment in Korea. A few cases of concentration averaging based on the application of CA BTP to domestic radioactive waste are presented, and the feasibility of the application is assessed. The radioactive waste considered herein does not satisfy the Disposal Concentration Limit (DCL) of the second-phase disposal facility while applying the preliminary classification. However, if CA BTP is applied when the radioactive waste is mixed with other radioactive waste items in a large and heavy container, it can be disposed of at the second-phase disposal facility in Gyeongju Repository. To apply the CA BTP of the U.S. NRC, it is necessary to investigate the safety assessment conditions of the US and Korea.