• 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.

• 대한전기학회:학술대회논문집
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• 대한전기학회 2009년도 제40회 하계학술대회
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• pp.662_663
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• 2009

Motor bus transfer involves the process of transferring a bus that has several critical motors to an alternate source of power when the main normal power source feeding them is interrupted. Bus transfer is a time-critical application in which the transfer progress depends on various parameters such as the type of motor, load on the motor at the time of transfer, inertia of the motor, and the combined open-circuit time constant of various motors present on the bus at the time of transfer. This paper present the result of modeling and simulation of nuclear power motor bus using ETAP(Electrical Transient Analyzing Program) program for motor and motor bus residual voltage decay characteristics.

• Nuclear Engineering and Technology
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• 제51권8호
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• pp.2026-2033
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• 2019

Fluidelastic instability of a tube array is a key factor of the security of a nuclear power plant. An unsteady model of the fluidelastic instability of a tube array subjected to two-phase flow was developed to analyze the fluidelastic instability of tube bundles in two-phase flow. Based on this model, a computational program was written to calculate the eigenvalue and the critical velocity of the fluidelastic instability. The unsteady model and the program were verified by comparing with the experimental results reported previously. The influences of void fraction and the tube's material properties on the critical velocity were investigated. Numerical results showed that, with increasing the void fraction of the two-phase flow, the tube array becomes more stable. The results indicate that the critical velocities of the tube array made of stainless are much higher than those of the other two tube arrays within void fraction ranging from 20% to 80%.

• 비파괴검사학회지
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• 제20권3호
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• pp.200-205
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• 2000

제 1보에서 Windows 환경 하에서 PSI/ISI 데이터의 효율적 관리 및 다원적평가와 분석이 가능한 고리 원전 PSI/ISI 데이터베이스 구축과 프로그램 개발(IDPIN)에 대해 소개하였다. 제 2보에서는 월성 원자력발전소 PSI/ISI 자료 관리를 위한 지능형 데이터베이스 프로그램(WS-IDPIN)을 개발하였다. 이 프로그램에는 월성 원전 PSI/ISI 관련 자료 분석 기능, PSI/ISI UT 검사 보고서 양식의 표준화와 전산 시스템화, PSI/ISI 비파괴검사 결과의 통계적 신뢰도 평가 프로그램(depth and length sizing performance 등) 등을 개발하여 원전 PSI/ISI 데이터 관리 전문가시스템(expert system)화에 한 단계 접근하였다.

• 방사성폐기물학회지
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• 제19권2호
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• pp.177-186
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• 2021

Radioactive materials from nuclear power facilities can be released into the atmosphere through various channels. Recently, the dispersion of radioactive materials has become critical issue in Korea after Kori Unit 1 and Wolsong Unit 1 were permanently shut down. In this study, annual atmospheric dispersion factors were compared based on the continuous release and purge release using the XOQDOQ computer program, a method for calculating atmospheric dispersion factors at commercial nuclear power stations. The meteorological data analyzed in this study was based on the Shin Kori nuclear power meteorological tower which has the largest operating nuclear power plants in Korea, for three years (from 2008 to 2010). The analysis results of the dispersion factor of the radioactive material release obtained using the XOQDOQ program showed that the difference between the continuous release and purge release was within two times. This study will be valuable helpful for revealing the uncertainty of the predictive atmospheric dispersion factor to achieve regulation.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2003년도 춘계학술대회
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• pp.610-616
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• 2003

Recently, the common concern of nuclear power industry in the development of technology mitigating and preventing the aging of steam generator tubes prevails, because the trends of steam generator flaws at Uljin unit #1,2 and KSNP(Korea Standard Nuclear Power Plant) impose a burden on the operation of nuclear power plant. While the regulatory agency is demanding the establishment of the advanced general performance maintenance system, the steam generator management program adapting advanced technology is being developed which may comply with EPRI PWR SG Guidelines based on NEI 97-06 ‘ General Guidelines including all the maintenance aspects consist of the tube integrity assessment criteria, repair limit, allowable leakage level, water chemistry will be composed in order to obtain the approval of regulatory agency and be applied to Nuclear power plant early 2005. This presentation is to introduce maintenance state including SG tube degradation and main contents of advanced SG management program being developed, and futhermore update present and future plan, and estimate the alternation after the completion.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 춘계학술발표회논문집(4)
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• pp.46-50
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• 1996

Tritium dose management is an important aspect of the radiation protection program at CANDU type reactor sites. This paper describes the bioassay and dosimetry of tritium at CANDU reactor sites, especially for Wolsung Nuclear Power Plant. It presents a compilation of information drawn from published papers, technical reports, international and national guidelines as well as practical experience both in Korean and Canadian CANDU Nuclear Power Plants. The implementation of this program would provide a technical basis for demonstrating to workers, managers and regulators that tritium bioassay measurements, dose calculations and records should be of acceptable quality and should meet overall radiation protection program objectives.

• 에너지공학
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• 제23권2호
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• pp.191-198
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• 2014

국내 가동중 원전은 발전소 안전성 및 신뢰성을 향상시킬 목적으로 정비효과감시 프로그램(정비규정)을 개발하여 2009년부터 이행 중에 있다. 정비효과감시 프로그램은 발전소 전 계통을 기능단위로 분류한 후 관리대상 선정, 안전중요도 결정 및 성능기준 수립을 통해 발전설비의 정비효과를 감시하는 프로그램이며, NUMARC 93-01 방법론을 기반으로 발전소별 고유 설계특성을 반영하여 개발하고 있다. 최근 건설 중인 APR1400형 원전도 운영초기 단계부터 정비체계 정착을 목적으로 안전관련계통 정비효과감시 프로그램을 개발하였으며, 향후 초기 성능평가를 통해 프로그램의 적합성을 검증할 예정이다. 결과적으로 고유 설계특성을 반영하여 개발한 정비효과감시 프로그램의 이행을 통해 APR1400형 원전의 안전성 및 신뢰성이 향상될 것으로 기대된다.

• Nuclear Engineering and Technology
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• 제51권6호
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• pp.1540-1553
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• 2019

A nuclear power plant (NPP) is a highly complex system-of-systems as manifested through its internal systems interdependence. The negative impact of such interdependence was demonstrated through the 2011 Fukushima Daiichi nuclear disaster. As such, there is a critical need for new strategies to overcome the limitations of current risk assessment techniques (e.g. the use of static event and fault tree schemes), particularly through simulation of the nonlinear dynamic feedback mechanisms between the different NPP systems/components. As the first and key step towards developing an integrated NPP dynamic probabilistic risk assessment platform that can account for such feedback mechanisms, the current study adopts a system dynamics simulation approach to model the thermal dynamic processes in: the reactor core; the secondary coolant system; and the pressurized water reactor. The reactor core and secondary coolant system parameters used to develop system dynamics models are based on those of the Palo Verde Nuclear Generating Station. These three system dynamics models are subsequently validated, using results from published work, under different system perturbations including the change in reactivity, the steam valve coefficient, the primary coolant flow, and others. Moving forward, the developed system dynamics models can be integrated with other interacting processes within a NPP to form the basis of a dynamic system-level (systemic) risk assessment tool.

• 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).