• 한국소음진동공학회:학술대회논문집
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• 한국소음진동공학회 2014년도 추계학술대회 논문집
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• pp.102-105
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• 2014

In accordance with U.S. Nuclear Regulatory Commission regulatory guide(NRC RG) 1.20(Rev.3), we are writing a comprehensive vibration assessment program(CVAP) design control document(DCD) and a technical report for U.S. NRC design certification(DC) of an Advanced Power Reactor 1400(APR1400) nuclear power plant(NPP). CVAP of an APR1400 NPP for U.S. NRC DC is classified as a non-prototype category 1 type. Therefore, CVAP DCD of reactor vessel internals(RVI) and steam generator internals(SGI) consist of analysis and full inspection program. However, piping system of primary and secondary system will be described as measurement program.

• 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.

• 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).

• 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.

• 한국소음진동공학회논문집
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• 제23권1호
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• pp.49-55
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• 2013

A reactor vessel internals comprehensive vibration assessment program(RVI CVAP) of an advanced power reactor 1400(APR1400) is being performed as a non-prototype category-2 type of reactor based on the US nuclear regulatory commission regulatory guide(NRC RG) 1.20. The aim of this paper is to present the results of structural response analysis and measuring locations of a upper guide structure(UGS) assembly of the APR1400 reactor. The analysis results of the UGS assembly show that the specified integrity levels meet the design acceptance criteria. Also, the measuring locations are determined by the analysis results of the UGS assembly and selection criteria of previous study. These analysis results and measuring locations will be used as a guide to design a measurement system for the APR1400 RVI CVAP.

• Nuclear Engineering and Technology
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• 제44권7호
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• pp.807-816
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• 2012

The Three Mile Island Unit 2 (TMI-2) accident has been studied extensively, as part of both post-accident technical assessment and follow-up computer code calculations. The models used in computer codes for severe accidents have improved significantly over the years due to better understanding. It was decided to reanalyze the severe accident scenario using current state of the art codes and methodologies. This reanalysis was adopted as a part of the joint standard problem exercise for the Atomic Energy Regulatory Board (AERB) - United States Regulatory Commission (USNRC) bilateral safety meet. The accident scenario was divided into four phases for analysis viz., Phase 1 covers from the accident initiation to the shutdown of the last Reactor Coolant Pumps (RCPs) (0 to 100 min), Phase 2 covers initial fuel heat up and core degradation (100 to 174 min), Phase 3 is the period of recovery of the core water level by operating the reactor coolant pump, and the core reheat that followed (174 to 200 min) and Phase 4 covers refilling of the core by high pressure injection (200 to 300 min). The base case analysis was carried out for all four phases. The majority of the predicted parameters are in good agreement with the observed data. However, some parameters have significant deviations compared to the observed data. These discrepancies have arisen from uncertainties in boundary conditions, such as makeup flow, flow during the RCP 2B transient (Phase 3), models used in the code, the adopted nodalisation schemes, etc. In view of this, uncertainty and sensitivity analyses are carried out using simulation based techniques. The paper deals with uncertainty and sensitivity analyses carried out for the first three phases of the accident scenario.

• 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.

• 대한전기학회:학술대회논문집
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• 대한전기학회 2011년도 제42회 하계학술대회
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• pp.2061-2062
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• 2011

국내 원자력발전소에 설치되어 있는 고압케이블은 장기운전에 따라 열화진단 및 유지보수의 필요성이 증대되고 있다. 주기적 안전성 평가(Periodic Safety Review) 및 계속운전과 관련하여서도 고압케이블의 관리방안이 지속적으로 요구되고 있다. 본 논문에서는 원자력발전소에 설치되어 있는 13.8kV급 고압케이블의 건전성을 확인하기 위하여 원자력발전소 계획예방정비 중 열화진단을 수행한 결과를 소개하고 있다. 특히 미국 NRC(Nuclear Regulatory Commission)에서는 케이블 진단 기법 11가지를 권고하고 있는데 본 논문에서는 그 중 국내 원자력발전소에 적용이 가능한 최신 케이블 진단 기술인 HFCT(High Frequency Current Transformer)를 이용한 부분방전(Partial Discharge ; PD) 측정, VLF(Very Low Frequency) 장치를 이용한 $Tan{\delta}$ 측정 및 열화상카메라를 이용한 케이블 온도 측정을 사용하여 고압 케이블 진단을 실시하였고, 그 결과를 서술하였다.

• Nuclear Engineering and Technology
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• 제50권2호
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• pp.306-312
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• 2018

There is an increasing need in the United States and around the world to move used nuclear fuel from wet storage in fuel pools to dry storage in casks stored at independent spent fuel storage installations or interim storage sites. Under normal conditions, the Nuclear Regulatory Commission limits cladding temperature to $400^{\circ}C$ for high-burnup (>45 GWd/mtU) fuel, with higher temperatures allowed for low-burnup fuel. An analysis was conducted with FRAPCON-4.0 on three modern fuel designs with three representative used nuclear fuel storage temperature profiles that peaked at $400^{\circ}C$. Results were representative of the majority of US light water reactor fuel. They conservatively showed that hoop stress remains below 90 MPa at the licensing temperature limit. Results also show that the limiting case for hoop stress may not be at the highest rod internal pressure in all cases but will be related to the axial temperature and oxidation profiles of the rods at the end of life and in storage.

• Nuclear Engineering and Technology
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• 제45권3호
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• pp.295-310
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• 2013

Industry- or regulatory-sponsored research activities on the resolution of Generic Safety Issue (GSI)-191 were reviewed, especially on the chemical effects. Potential chemical effects on the head loss across the debris-loaded sump strainer under a post-accident condition were experimentally evidenced by small-scale bench tests, integrated chemical effects test (ICET), and vertical loop head loss tests. Three main chemical precipitates were identified by WCAP-16530-NP: calcium phosphate, aluminum oxyhydroxide, and sodium aluminum silicate. The former two precipitates were also identified as major chemical precipitates by the ICETs. The assumption that all released calcium would form precipitates is reasonable. CalSil insulation needs to be minimized especially in a plant using trisodium phosphate buffer. The assumption that all released aluminum would form precipitates appears highly conservative because ICETs and other studies suggest substantial solubility of aluminum at high temperature and inhibition of aluminum corrosion by silicate or phosphate. The industry-proposed chemical surrogates are quite effective in increasing the head loss across the debris-loaded bed and more effective than the prototypical aluminum hydroxide precipitates generated by in-situ aluminum corrosion. There appears to be some unresolved potential issues related to GSI-191 chemical effects as identified in NUREG/CR-6988. The United States Nuclear Regulatory Commission, however, concluded that the implications of these issues are either not generically significant or are appropriately addressed, although several issues associated with downstream in-vessel effects remain.