• 한국항해항만학회지
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• 제40권1호
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• pp.21-26
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• 2016

다른 응용분야에 비하여 개략적인 성능을 요구했던 해양 분야에서도 최근 들어 항만 영역에 따라 수치적으로 구체화된 성능을 요구하고 있고, 측위 정확도뿐만 아니라 무결성, 연속성, 가용성 면에서도 보장을 요구하고 있다. 현 DGPS 기준국의 서비스를 사용하는 사용자 입장에서, DGPS 서비스는 아직 IMO에서 요구하는 성능을 모두 만족시키기에 불충분하다. 특히, HEA 영역에서 요구하는 측위 정확도 수준은 대부분 만족하나 무결성 측면에서는 성능을 보장할 수 없다. 본 논문에서는 기존 DGPS 기준국의 감시, 보장, 공지 기능을 강화하여 무결성 측면에서 보강하기 위한 향상된 감시 개념을 제안하고자 한다. 이를 위하여 현재 DGPS 서비스 감시 기능의 취약점을 제시하고 이에 대한 영향을 분석하였으며 이를 토대로 대책을 마련하였다. 본 연구결과는 추후 DGPS 서비스 개선을 위한 기초연구로 활용할 수 있을 것이다.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2003년도 춘계학술발표대회 요약집
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• pp.146-146
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• 2003

• 한국압력기기공학회 논문집
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• 제5권2호
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• pp.13-19
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• 2009

As the operating time in nuclear power plants (NPPs) increases, the integrity of nuclear components may be continually degraded due to aging effects of systems, structures and components. Recently, a number of NPPs are being operated beyond their design life to produce more electricity without shutting down. The critical issue in extending a lifetime is to maintain the level of safety during the extended operation period while satisfying the international regulatory standards. Therefore, it is beneficial to build a monitoring system to measure an aging status. In this paper, the Aging Monitor (AM) based on lots of aging database obtained from the operating plants and research results on the aging effects was developed to monitor, manage and evaluate the aging phenomena systematically and effectively in NPPs. The AM for the CANDU is divided into 6 modules: (1) Aging Alarm/Coloring Monitor, (2) Aging Database, (3) Aging Document, (4) Real-time Integrity Monitor, (5) Surveillance and Inspection Management System, and (6) Continued Operation and Periodic Safety Review (PSR) Safety Evaluation. The proposed system is expected to provide the integrity assessment for the major mechanical components of an NPP under concurrent working environments.

• Nuclear Engineering and Technology
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• 제47권6호
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• pp.766-775
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• 2015

The integrity assessment of reactor vessel internals should be conducted in the design process to secure the safety of nuclear power plants. Various loads such as self-weight, seismic load, flow-induced load, and preload are applied to the internals. Therefore, the American Society of Mechanical Engineers (ASME) Code, Section III, defines the stress limit for reactor vessel internals. The present study focused on structural response analyses of the upper guide structure upper flange. The distributions of the stress intensity in the flange body were analyzed under various design load cases during normal operation. The allowable stress intensities along the expected sections of stress concentration were derived from the results of the finite element analysis for evaluating the structural integrity of the flange design. Furthermore, seismic analyses of the upper flange were performed to identify dynamic behavior with respect to the seismic and impact input. The mode superposition and full transient methods were used to perform time-history analyses, and the displacement at the lower end of the flange was obtained. The effect of the damping ratio on the response of the flange was also evaluated, and the acceleration was obtained. The results of elastic and seismic analyses in this study will be used as basic information to judge whether a flange design meets the acceptance criteria.

• 한국안전학회지
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• 제33권6호
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• pp.58-65
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• 2018

The life span of the porcelain insulator was made to be 30 years, but currently many of the 154kV NGK porcelain insulators using in Korea are found to have passed the production life. Accidents caused by aged mechanical breakdown can lead to disruption of power supply in some areas, large economic losses, and casualties. Therefore, ultrasonic method, which is one of the non - destructive test methods, is applied as a method for evaluating the integrity of porcelain insulators. In this study, the experiment on the interface of cap was conducted and the difference between the energy difference and the attenuation coefficient of the reflected wave was derived according to the interface state of the steel - cement. The results of this study are expected to be used as the basic data of the ultrasonic testing to evaluate the interface condition of the porcelain insulator cap.

• 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.

• 방사성폐기물학회지
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• 제18권spc호
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• pp.21-36
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• 2020

With respect to spent nuclear fuels, disposal containers and bentonite buffer blocks in deep geological disposal systems are the primary engineered barrier elements that are required to isolate radioactive toxicity for a long period of time and delay the leakage of radio nuclides such that they do not affect human and natural environments. Therefore, the thermal stability of the bentonite buffer and structural integrity of the disposal container are essential factors for maintaining the safety of a deep geological disposal system. The most important requirement in the design of such a system involves ensuring that the temperature of the buffer does not exceed 100℃ because of the decay heat emitted from high-level wastes loaded in the disposal container. In addition, the disposal containers should maintain structural integrity under loads, such as hydraulic pressure, at an underground depth of 500 m and swelling pressure of the bentonite buffer. In this study, we analyzed the thermal stability and structural integrity in a deep geological disposal environment of the improved deep geological disposal systems for domestic light-water and heavy-water reactor types of spent nuclear fuels, which were considered to be subject to direct disposal. The results of the thermal stability and structural integrity assessments indicated that the improved disposal systems for each type of spent nuclear fuel satisfied the temperature limit requirement (< 100℃) of the disposal system, and the disposal containers were observed to maintain their integrity with a safety ratio of 2.0 or higher in the environment of deep disposal.

• Nuclear Engineering and Technology
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• 제54권5호
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• pp.1698-1711
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• 2022

A combined safety approach, which uses a best-estimate computer code and adopts conservative assumptions for safety systems availability, is developed and applied to the safety margin evaluation for the Loss of Condenser Vacuum (LOCV) of the 1000 MWe Korean Nuclear Power Plant. The Multi-dimensional Analysis of Reactor Safety-KINS standard (MARS-KS) code is selected as a best-estimate code and the PAPIRUS program is used to obtain different initial operational conditions through random sampling of control variables. During an LOCV event, fuel integrity is not threatened by the increase in Departure from Nuclear Boiling Ratio (DNBR). However, the high pressure in the primary coolant system and the secondary system might affect the system integrity. Thus, the peak pressure becomes a major safety concern. Transient analyses are performed for 124 cases of different initial conditions and the most conservative case, which results in the highest system pressure is selected. It is found the suggested methodology gives similar peak pressures when compared to those predicted from existing methodologies. The proposed approach is expected to minimize the time and efforts required to identify the conservative plant conditions in the existing conservative safety methodologies.

• 한국안전학회지
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• 제22권4호
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• pp.7-12
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• 2007

In order to evaluate the integrity of Zr-2.5Nb pressure tubes, probabilistic fracture mechanics(PFM) approach was employed. Failure assessment diagram(FAD), plastic collapses, and critical crack lengths(CCL) were used for evaluating the failure probability as failure criteria. The Kr-FAD as failure assessment diagram was used because fracture of pressure tubes occurred in brittle manner due to hydrogen embrittlement of material by deuterium fluence. The probabilistic integrity evaluation observed AECL procedures and used fracture toughness parameters of EPRI and recently announced theory. In conclusion, the probabilistic approach using the Kr-FAD made it possible to determine major failure criterion in the pressure tube integrity evaluation.

• 한국안전학회지
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• 제23권4호
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• pp.7-12
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• 2008

This study proposes integrity evaluation method of weld zone in titanium using titanium distance amplitude characteristics curve(TDACC) and ultrasonic signals. For these purposes, the ultrasonic signals for porosity defects of weld zone in titanium are acquired in the type of time series data and echo strength. 4 lines in the DACC indicated damage evaluation standard of weld zone in titanium. The acquired ultrasonic signals agree fairly well with the measured results of reference block and sensitivity block(defect location, bean1 propagation distance, echo strength, etc). The proposed TDACC in this study agree fairy well with the measured results of weld zone in titanium(weld defects as porosity). The proposed TDACC in this study can be used for integrity evaluation of weld zone in titanium.