• 대한전기학회:학술대회논문집
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• 대한전기학회 2005년도 제36회 하계학술대회 논문집 B
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• pp.1008-1010
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• 2005

In the examination of steam generator(SG) tube in nuclear power plant, eddy current testing probes play an important role in detecting the defects. Bobbin probe and MRPC probe is usually used for the inspection of SG tube. Bobbin probe is good at high speed inspection, but ability of detection of circumferential defect is very weak. On the contrary MRPC probe, which moves for inspection in the direction of axial and circumferential simultaneously, has very slow inspection speed, but it has excellent detection capability for small cracks, which is hardly detected by bobbin probe. In this paper, for the accurate analysis of experimental ECT signals, construction of MRPC probe signals database according to the variation of defect size is the main purpose. Using 3-D finite element method, ECT signals are analyzed, and signals analysis add according to frequency ingredient. The results, which are analysis and characteristics ion of electromagnetism simulation signals, is databased.

• Nuclear Engineering and Technology
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• 제50권8호
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• pp.1306-1313
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• 2018

The randomness and incipient nature of certain faults in reactor systems warrant a robust and dynamic detection mechanism. Existing models and methods for fault diagnosis using different mathematical/statistical inferences lack incipient and novel faults detection capability. To this end, we propose a fault diagnosis method that utilizes the flexibility of data-driven Support Vector Machine (SVM) for component-level fault diagnosis. The technique integrates separately-built, separately-trained, specialized SVM modules capable of component-level fault diagnosis into a coherent intelligent system, with each SVM module monitoring sub-units of the reactor coolant system. To evaluate the model, marginal faults selected from the failure mode and effect analysis (FMEA) are simulated in the steam generator and pressure boundary of the Chinese CNP300 PWR (Qinshan I NPP) reactor coolant system, using a best-estimate thermal-hydraulic code, RELAP5/SCDAP Mod4.0. Multiclass SVM model is trained with component level parameters that represent the steady state and selected faults in the components. For optimization purposes, we considered and compared the performances of different multiclass models in MATLAB, using different coding matrices, as well as different kernel functions on the representative data derived from the simulation of Qinshan I NPP. An optimum predictive model - the Error Correcting Output Code (ECOC) with TenaryComplete coding matrix - was obtained from experiments, and utilized to diagnose the incipient faults. Some of the important diagnostic results and heuristic model evaluation methods are presented in this paper.

• Nuclear Engineering and Technology
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• 제53권11호
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• pp.3580-3596
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• 2021

An integral test facility, PILLAR, was commissioned, aiming to provide valuable experimental results which can be referenced by system and component designers and used for the performance demonstration of liquid-metal-cooled, passive small modular reactors (SMRs) toward their licensing. The setup was conceptualized by a scaling analysis which allows the vertical arrangements to be conserved from its prototypic reactor, scaled uniformly in the radial direction achieving a flow area reduction of 1/200. Its final design includes several heater rods which simulate the reactor core, and a single heat exchanger representing the steam generators in the prototype. The system behaviors were characterized by its data acquisition system implementing various instruments. In this paper, we present not only a detailed description of the facility components, but also selected experimental results of both steady-state and transient cases. The obtained steady-state test results were utilized for the benchmark of a system code, achieving a capability of accurate simulations with ±3% of maximum deviations. It was followed by qualitative comparisons on the transient test results which indicate that the integral system behaviors in passive LBE-cooled systems are able to be predicted by the code.

• Nuclear Engineering and Technology
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• 제54권2호
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• pp.554-564
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• 2022

The natural circulation phenomena occurring in fully integrated nuclear reactors are associated with a unique formation mechanism. The phenomenon results from a structural feature of these reactors involving upward flow from the core, located in the central-bottom region of a single vessel, and downward flow to the steam generator in the annulus region. In this study, to understand the natural circulation in a single vessel involving a multi-layered flow path, single-phase and two-phase natural circulation tests were performed using the SMART-ITL facility, and validation analysis of the TASS/SMR-S code was performed by comparing the corresponding test results. Three single-phase natural circulation tests were sequentially conducted at 15%, 10%, and 5% of full-scaled core-power without RCP operation, following which a two-phase natural circulation test was successively conducted with an artificial discharge of coolant inventory. The simulation capability of the TASS/SMR-S code with respect to the natural circulation phenomena was validated against the test results, and somewhat conservative but reasonably comparative results in terms of overall thermalhydraulic behavior were shown.

• 에너지공학
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• 제13권1호
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• pp.51-59
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• 2004

국내에 설치 운영중인 원전 훈련용 시뮬레이터의 핵 증기공급 계통 열수력 프로그램은 1980련 전후에 외국 벤더들이 개발하여 공급한 것으로 이들 열수력 프로그램은 핵 증기공급 계통 열수력 현상을 실시간으로 모의하기 위해 과도하게 단순화된 모델을 채택하고 있다. 그 결과 원자로 냉각계통에 복잡한 이상유동이 발생하는 사고를 모의하는 경우 정확도가 떨어질 수 있어 부정적인 훈련(Negative training)을 초래할 가능성이 있다. 이와같은 문제를 해결하기 위해 전력연구원에서는 RETRAN-3D코드를 기본으로 시뮬레이터용 핵 증기공급 계통 열수력 프로그램 ARTS코드를 개발하였다. RETRAN-3D코드를 기본으로 하는 ARTS코드는 거의 대부분의 사고를 실시간으로 모의할 수 있으며 계산의 건전성도 보장된다. 그러나, 대형냉각재 상실사고나 저압 저유속 상태의 장기 과도현상 등을 모의하는 경우에 발생하는 계산실패나 실시간 계산 지체등의 가능성이 있다. 이 경우 이를 자동으로 대체 보완할 수 있는 보조계산체계를 개발했다. 특히, ARTS코드의 실시간 계산 및 건전성 문제가 예상되는 대형냉각재 상실사고를 주모의 대상으로 간주했다. 계산 결과는 코드의 정확도, 실시간 계산능력, 건전성 및 운전원 교육등에서 최종안정성평가보고서 및 ANSI/ANS-3.5-1998$^{[1]}$ 시뮬레이터 소프트웨어 기준을 만족하는 것으로 평가되었다

• 융합정보논문지
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• 제10권10호
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• pp.183-195
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• 2020

본 연구에서는 대학수학 메이커수업 특성의 요인인 강사역량, 교육프로그램, 교육서비스, 물리적 교육환경 요인과 대학생들의 수학에 대한 흥미 및 태도 간의 영향관계를 검증하는 것을 목적으로 하였다. 부산권 소재 대학교에 재학 중인 이공계열 남녀 대학생 228명을 대상으로 설문조사를 실시하였고, SPSS 26.0 프로그램을 활용하여 실증분석을 수행하였다. 연구결과 첫째, 대학수학 메이커수업 특성 요인 중 강사역량(β=.349, t=6.380, p<.001), 교육프로그램(β=.361, t=5.650, p<.001), 물리적 교육환경(β=.196, t=3.281, p<.01) 요인이 대학생의 수학에 대한 흥미에 유의미한 정(+)의 영향을 미쳤다. 둘째, 대학수학 메이커수업에 대한 흥미(β=.349, t=6.380, p<.001) 요인은 대학생의 수학에 대한 태도에 유의미한 정(+)의 영향을 미치는 것으로 나타났다. 셋째, 대학수학 메이커수업 특성 요인 중 강사역량(β=.340, t=6.365, p<.001), 교육프로그램(β=.352, t=5.559, p<.001), 물리적 교육환경(β=.226, t=3.537, p<.01) 요인이 대학생의 수학에 대한 태도에 유의미한 정(+)의 영향을 미치는 것으로 나타났다. 본 연구의 결과를 통해 대학수학 메이커수업의 교육프로그램 수준과 강사역량이 높고, 물리적 교육환경이 뛰어날 때, 대학생의 수학에 대한 태도 뿐 아니라 궁극적으로 수학에 대한 태도에도 긍정적인 영향을 미칠 수 있다는 결론에 도달하였다.

• 대한기계학회논문집A
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• 제40권1호
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• pp.81-86
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• 2016

밸브는 배관의 유량을 차단 및 제어하기 위한 장치로써 게이트 밸브, 글로브 밸브, 체크 밸브 등 많은 종류가 사용되고 있다. 그 중 글로브 밸브는 고압력 조건에서의 유량조절이 용이하여 LNG 선박, 증기 배관 등에 사용된다. 본 논문에서는 글로브 밸브의 누설 문제를 구조적으로 해결하기 위해 시트의 형상을 변형하는 방법을 제시하였다. 또한 유한 요소 해석을 통해 각 모델의 응력분포와 변형량을 비교하고 이를 통하여 제안한 모델에 대한 검증을 진행하였다. 시뮬레이션 결과 제안된 모델에서 원주 방향의 변형이 줄어들고, 누설을 감소시킬 수 있는 Self-supporting 효과를 확인할 수 있었다.

• 비파괴검사학회지
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• 제20권1호
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• pp.1-9
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• 2000

본 연구에서는 유도초음파를 이용하여 열 교환기와 증기발생기 튜브의 결함을 비파괴적으로 탐상하고 그 크기를 산정하였다. 이론적인 해석을 위해 인코넬 (Inconel) 튜브에 대한 위상 및 군속도 분산선도를 Longitudinal 모드와 Flexural 모드에 대해 구하였다. 튜브의 원주방향 레이저노치와 튜브 지지대 하단의 방전가공결함(EDM wear)을 각각 비대칭 및 대칭 탐촉자 세트를 사용하여 탐상하였다. 실험결과 방전가공결함은 L(0, 2), L(0, 3), L(0, 4) 모드로 탐상되었으며, 그 중 L(0, 4) 모드가 결함으로부터 가장 잘 반사되었다. 레이저노치의 경우에는 L(0, 1) 모드 주변의 Flexural 모드가 결함을 탐상하고 크기를 산정하는데 사용될 수 있음을 보였다.

• Nuclear Engineering and Technology
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• 제41권8호
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• pp.1053-1064
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• 2009

The existence of a large sodium pool in the KALIMER, a pool-type LMR developed by the Korea Atomic Energy Research Institute, plays an important role in reactor safety and operability because it determines the grace time for operators to cope with an abnormal event and to terminate a transient before reactor enters into an accident condition. A two-dimensional hot pool model has been developed and implemented in the SSC-K code, and has been successfully applied for the assessment of safety issues in the conceptual design of KALIMER and for the analysis of anticipated system transients. The other important models of the SSC-K code include a three-dimensional core thermal-hydraulic model, a reactivity model, a passive decay heat removal system model, and an intermediate heat transport system and steam generation system model. The capability of the developed two-dimensional hot pool model was evaluated with a comparison of the temperature distribution calculated with the CFX code. The predicted hot pool coolant temperature distributions obtained with the two-dimensional hot pool model agreed well with those predicted with the CFX code. Variations in the temperature distribution of the hot pool affect the reactivity feedback due to an expansion of the control rod drive line (CRDL) immersed in the pool. The existing CRDL reactivity model of the SSC-K code has been modified based on the detailed hot pool temperature distribution obtained with the two-dimensional pool model. An analysis of an unprotected transient over power with the modified reactivity model showed an improved negative reactivity feedback effect.

• Nuclear Engineering and Technology
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• 제47권2호
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• pp.204-211
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• 2015

Probabilistic safety assessment (PSA) has had a significant role in quantitative decision-making by finding design and operational vulnerabilities and evaluating cost-benefit in improving such weak points. In particular, it has been widely used as the core methodology for risk-informed applications (RIAs). Even though the nature of PSA seeks realistic results, there are still "conservative" aspects. One of the sources for the conservatism is the assumptions of safety analysis and the estimation of failure frequency. Surveillance, diagnosis, and prognosis (SDP), utilizing massive databases and information technology, is worth highlighting in terms of its capability for alleviating the conservatism in conventional PSA. This article provides enabling techniques to solidify a method to provide time- and condition-dependent risks by integrating a conventional PSA model with condition monitoring and prognostics techniques. We will discuss how to integrate the results with frequency of initiating events (IEs) and probability of basic events (BEs). Two illustrative examples will be introduced: (1) how the failure probability of a passive system can be evaluated under different plant conditions and (2) how the IE frequency for a steam generator tube rupture (SGTR) can be updated in terms of operating time. We expect that the proposed model can take a role of annunciator to show the variation of core damage frequency (CDF) depending on operational conditions.