• 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.

• 동력기계공학회지
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• 제10권4호
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• pp.71-77
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• 2006

This study introduces the seismic qualification of safety related equipments for nuclear power plants to verify the possibility of resonance in regard to the operating speed and the structural integrity due to external piping nozzle loads as well as seismic dynamic loads using El-Centro earthquake, which was occurred in the 1940's previously. As a first step, it is necessary to investigate the natural frequency of the vertical mixed flow pump in order to determine whether static or dynamic equipment comparing with seismic cut-off frequency, 33hz. Also the normal mode analysis was carried out with the introduction of seismic redesign straint at the middle of vertical pump to increase the natural frequency. In terms of structural integrity, the application of static analysis with normal, upset and faulted nozzle loads event was presented for the comparison of material allowable stress. Also the dynamic analysis was performed to show the design adequacy through the application to the case of El-Centro earthquake.

• 대한용접접합학회:학술대회논문집
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• 대한용접접합학회 2008년도 춘계학술발표대회 개요집
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• pp.9-9
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• 2008

• 대한기계학회논문집A
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• 제29권5호
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• pp.734-742
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• 2005

The present study proposes the integrity evaluation model for a straight pipe with local wall thinning defect, which reflects the characteristics of training shape and loading condition in the Piping of nuclear power plant. For this purpose, a series of finite element analyses are performed under various defect geometries and loading conditions, and real pipe experiment data performed previously is employed. The model includes the effect of thinning length as well as thinning depth and width, and also it considers the combined loading effect between internal pressure and bending moment. The proposed model has been validated using the results of finite element analysis and pipe experiment data. The results indicate that the proposed model provides more reliable predictions of pipe failure than the current existing model, in terms of accuracy, consistency, and conservativeness of results.

• 한국건축시공학회:학술대회논문집
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• 한국건축시공학회 2016년도 춘계 학술논문 발표대회
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• pp.91-92
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• 2016

Structural integrity of reinforced concrete structures including nuclear power plants needs to be evaluated on a regular basis. Deterioration inside the concrete structures can be represented by voids. In this study, the varied volume fraction of voids inside mortar specimens was studied as a parameter using ultrasonic testing equipments. Both direct and indirect measurement methods were employed. The results show that there is a clear distinction between the specimens with different void volume fractions.

• Structural Engineering and Mechanics
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• 제37권3호
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• pp.257-265
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• 2011

Plant-specific analyses of 5 types of domestic reactors in Korea are performed to assure the structural integrity of the reactor pressure vessel (RPV) during transients which are expected to initiate pressurized thermal shock (PTS) events. The failure probability of the RPV due to PTS is obtained by performing probabilistic fracture mechanics analysis. The through-wall cracking frequency is calculated and compared to the acceptance criterion. Considering the fluence at the end of life expected by surveillance test, the sufficient safety margin is expected for the structural integrity of all reactor pressure vessels except for the oldest one during the pressurized thermal shock events. If the flaw with aspect ratio of 1/12 is considered to eliminate the conservatism, the acceptance criteria is not exceeded for all plants until the fluence level of $8{\times}10^{19}\;n/cm^2$, generating sufficient margin beyond the design life.

• KSII Transactions on Internet and Information Systems (TIIS)
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• 제12권12호
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• pp.6079-6097
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• 2018

Safety-related systems (SRSs) has widely used in shipbuilding and power generation to prevent fatal accidents and to protect life and property. Thus, SRS performance is a high priority. The safety integrity level (SIL) is the relative performance level of an SRS with regard to its ability to operate reliably in a safe manner. In this article, we proposed an optimal design procedure to achieve the targeted SIL of SRSs. In addition, a more efficient failure mode and effects diagnostic analysis (FMEDA) process and optimization model were developed to improve cost efficiency. Based on previous IEC 61508 diagnostic analyses that revealed unnecessary costs associated with excessive reliability, the new approach consists of two phases: (i) SIL evaluation by FMEDA, and (ii) solution optimization for achieving the target SIL with minimal cost using integer-programming models. The proposed procedure meets the required safety level and minimizes system costs. A case study involving a gas-detection SRS was conducted to demonstrate the effectiveness of the new procedure.

• 한국기계가공학회지
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• 제18권5호
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• pp.53-59
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• 2019

Continuous ship unloaders (CSUs) are used for the uninterrupted transport of material in processing industries, power plants, and harbors in accordance with the stream rate of the material. This study analyzed the structural integrity and fatigue life of a CSU structure using finite element structural analysis in ANSYS APDL software. The stress varied greatly depending on the luffing angle and the slew angle of the boom conveyor. The structural integrity of the CSU girder was evaluated by applying ASME BPVC Section VIII Division 2. The fatigue cycle at the angle with the greatest stress difference was calculated. The fatigue cycle was calculated by applying the JIS B 8821:2013 fatigue curve. It was confirmed that the fatigue cycle of the CSU satisfies the allowable fatigue of 200,000 cycles.

• KSII Transactions on Internet and Information Systems (TIIS)
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• 제17권3호
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• pp.938-957
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• 2023

The limited computing power of sensor nodes in wireless sensor networks (WSNs) and data tampering during wireless transmission are two important issues. In this paper, we propose a scheme for independent individual authentication of WSNs data based on digital watermarking technology. Digital watermarking suits well for WSNs, owing to its lower computational cost. The proposed scheme uses independent individual to generate a digital watermark and embeds the watermark in current data item. Moreover, a sink node extracts the watermark in single data and compares it with the generated watermark, thereby achieving integrity verification of data. Inherently, individual validation differs from the grouping-level validation, and avoids the lack of grouping robustness. The improved performance of individual integrity verification based on proposed scheme is validated through experimental analysis. Lastly, compared to other state-of-the-art schemes, our proposed scheme significantly reduces the false negative rate by an average of 5%, the false positive rate by an average of 80% of data verification, and increases the correct verification rate by 50% on average.

• 한국안전학회지
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• 제29권5호
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• pp.160-165
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• 2014

The realistic cases causing severe core damage should be analyzed and arranged systematically for preparing an accident management of the specific nuclear power plant. The objective of this paper is to establish basic technical information for reactor safety and reactor building integrity management strategies in CANDU reactor severe accident. For the development of severe accident management strategies, plant specific features and behaviors must be studied by detailed analysis works. This analysis scope will serve to cover overall methods and analyzing results to understand the reactor building integrity status in the most likely severe accident sequences that could occur at CANDU reactor. Also analysis results could help prevent or mitigate severe accidents for the identification of any plant specific vulnerabilities to severe accidents using the probabilistic safety assessment (PSA) quantified results.