• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.13 no.1
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• pp.1-10
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• 2015

In the safety assessment of the geological disposal system of the radioactive wastes, the abnormal scenarios, in which the system is impacted by the abnormal events, need to be considered in addition to the reference scenario. In this study, characterization and prediction of well intrusion as one of the abnormal events which will impact the disposal system were conducted probabilistically and statistically for the safety assessment. The domestic well development data were analyzed, and the prediction methodologies of the well intrusion were suggested with a computation example. From the results, the annual well development rate per unit area in Korea was about 0.8 well/yr/km² in the conservative point of view. Considering the area of the overall disposal system which is about 1.5 km², the annual well development rate within the disposal system could be 1.2 well/yr. That is, it could be expected that more than one well would be installed within the disposal system every year after the institutional management period. From the statistical analysis, the probabilistic distribution of the well depth followed the log-normal distribution with 3.0363 m of mean value and 1.1467 m of standard deviation. This study will be followed by the study about the impacts of the well intrusion on the geological disposal system, and the both studies will contribute to the increased reliability of safety assessment.

• The Transactions of the Korea Information Processing Society
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• v.13 no.5
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• pp.227-235
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• 2024

Since nuclear power plants (NPPs) increasingly employ digital I&C systems, reliability evaluation for NPP software has become crucial for NPP probabilistic risk assessment. Several methods for estimating software reliability have been proposed, but there is no available tool support for those methods. To support NPP software manufacturers, we propose a reliability measurement tool for NPP software. We designed our tool to provide reliability estimation depending on available qualitative and quantitative information that users can offer. We applied the proposed tool to an industrial reactor protection system to evaluate the functionality of this tool. This tool can considerably facilitate the reliability assessment of NPP software.

• KSCE Journal of Civil and Environmental Engineering Research
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• v.40 no.5
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• pp.455-463
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• 2020

Road facilities with a service life of more than 30 years are expected to triple in the next ten years. The seismic performance of road facilities should be reviewed with consideration of the "Common Application of Seismic Design Standards" issued by Korea's Ministry of Public Administration and Security in 2017. These standards should be applied to all existing road facilities, including retrofitted or seismic-designed facilities, for evaluating seismic performance. In order to manage seismic performance for a large number of facilities, decision-support technology that can provide economic and reliable results is needed. However, the indices method currently used in Korea is a deterministic method, and the seismic performance of individual facilities is evaluated based on qualitative indices so that only retrofitting among road facilities is prioritized. In turn, with the indices method, it is difficult to support decisions other than the decision to prioritize retrofitting. Therefore, it is necessary to use the seismic risk assessment method to overcome such shortcomings and provide useful information such as direct loss, indirect socio-economic loss, and benefit of the investment.

• Journal of Radiation Protection and Research
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• v.40 no.4
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• pp.244-251
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• 2015

The containment filtered venting system reduces the range of the contamination area around the nuclear power plant by strengthening the integrity of the containment building. In this study, the probabilistic assessment code MACCS2 was used to assess the effect of the CFVS to off-site. The accident source term was selected from a Probabilistic Safety Analysis report of SHINKORI 1&2 Nuclear Power Plant. The three source term categories from 19 STC were chosen to evaluate the effective dose and thyroid dose of residents around the power plant and the dose with CFVS and without CFVS were compared. The dose was calculated according to the distance from the nuclear power plant, so the damage scale based on the distance that exceeds the IAEA criteria for effective dose (100 mSv per 7 days) and thyroid dose (50 mSv per 7 days) were compared. The effective dose reduction rates of the STC-3, STC-4, STC-6 were about 95-99% in the whole range (0~35 km), 96-98% for the thyroid dose. There are similar results between effective dose and thyroid dose. After applying the CFVS, the damage scale that exceeds the effective dose criteria was about 1 km (mean). Especially, the STC-4 damage scale was decreased from 26 km (mean) to 1.2 km (mean) significantly. The damage scale that exceed the thyroid dose criteria was decreased to 2~3 km (mean). The STC-4 damage scale was also decreased significantly as compared to STC-3, STC-6 in terms of effective dose.

• Journal of the Korean Society of Safety
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• v.34 no.3
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• pp.102-110
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• 2019

This paper describes the work and results of the preliminary Probabilistic Safety Assessment (PSA) for a research reactor in the design phase. This preliminary PSA was undertaken to assess the level of safety for the design of a research reactor and to evaluate whether it is probabilistically safe to operate and reliable to use. The scope of the PSA described here is a Level 1 PSA which addresses the risks associated with core damage. After reviewing the documents and its conceptual design, eight typical initiating events are selected regarding internal events during the normal operation of the reactor. Simple fault tree models for the PSA are developed instead of the detailed model at this conceptual design stage. A total of 32 core damage accident sequences for an internal event analysis were identified and quantified using the AIMS-PSA. LOCA-I has a dominant contribution to the total CDF by a single initiating event. The CDF from the internal events of a research reactor is estimated to be 7.38E-07/year. The CDF for the representative initiating events is less than 1.0E-6/year even though conservative assumptions are used in reliability data. The conceptual design of the research reactor is designed to be sufficiently safe from the viewpoint of safety.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.15 no.1
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• pp.33-39
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• 2019

An 1,500MW advanced power reactor required the standard design approval by a Korean regulatory body in 2014. The reactor has been designed to have a 4-train independent safety concept and a passive auxiliary feedwater system (PAFS). The full power risk or core damage frequency (CDF) of 1,500MW advanced power reactor has been reduced more than that of APR1400. However, the risk during the low power and shutdown (LPSD) operation should be reduced because CDF of LPSD is about 4.7 times higher than that of internal full power. The purpose of paper is to analysis design alternatives to reduce risk during the LPSD. This paper suggests design alternatives to reduce risk and presents sensitivity analysis results.

• Journal of the Korean Geotechnical Society
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• v.22 no.6
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• pp.51-61
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• 2006

An efficient and accurate hybrid reliability analysis method is proposed in this paper to quantify the risk of an axially loaded single pile considering pile-soil interaction behavior and uncertainties in various design variables. The proposed method intelligently integrates the concepts of the response surface method, the finite difference method, the first-order reliability method, and the iterative linear interpolation scheme. The load transfer method is incorporated into the finite difference method for the deterministic analysis of a single pile-soil system. The uncertainties associated with load conditions, material and section properties of a pile and soil properties are explicitly considered. The risk corresponding to both serviceability limit state and strength limit state of the pile and soil is estimated. Applicability, accuracy and efficiency of the proposed method in the safety assessment of a realistic pile-soil system subjected to axial loads are verified by comparing it with the results of the Monte Carlo simulation technique.

• Nuclear Engineering and Technology
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• v.26 no.3
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• pp.355-366
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• 1994

The USNRC issued a revised ECCS rule that allows the use of best estimate computer codes for safety analysis. The rule also requires an estimation of uncertainty in calculated system response when applying the best estimate computer codes. A practical realistic evaluation methodology to evaluate the ECCS performance that satisfies the requirements of the ECCS rule has been developed and this paper describes the application of new realistic evaluation methodology to large break LOCA for, the demonstration of the new methodology. The computer code RELAP5/MOD3/KAERI, which was improved from RELAP5/MOD3.1, was used as the best estimate code in the application. The uncertainty of the code was evaluated by assessing several separate and integral effect tests, and for the application to actual plant Kori 3 & 4 was selected as the reference plant. Response surfaces for blowdown and reflood PCTs were generated from the results of the sensitivity analyses and probability distribution functions were established by random sampling or Monte-Carlo method for each response surface. Final uncertainties were quantified at 95% probability level and safety margins for large break LOCA were discussed.

• Journal of the Korean Society of Marine Environment & Safety
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• v.13 no.1 s.28
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• pp.55-60
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• 2007

Many electric power cables span the sea in Korean archipelago. Lots of shipping routes were established between the islands in the same area and ships frequently sail under the cable which cross the seas. Sometimes electric power cables were accidently broken by sailing ships and catastrophic damage of island industry followed thereby. If navigators have detailed knowledge about the height of electric power cable, the ship's sailing condition will be greatly improved. But at the present time, navigators have limited data about the electric power cable. Those are the horizontal distance between pylons, the height of pylons and minimum height of electric power cable. This study introduced the calculating methodology to find out the height of cable at any position between pylons. The ship's tracks were recorded and traffic safety was assessed by statistical method in relation to cable height.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05b
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• pp.570-575
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• 1996

1단계 확률론적 안전성 평가 (Level 1 Probabilistic Safety Assessment, PSA)를 수행할 때 나타나는 보조계통 고장 수목간의 순환 논리는 사고 경위 정량화를 위하여 해결되어야만 한다. 기존의 PSA에서는 이를 위하여 별도의 고장 수목을 다시 작성하였으나, 이 방법을 사용하기 위하여서는 보조계통 간의 관계를 검토하여야 하며, 이에 따른 별도의 고장 수목을 작성하여야 하는 등 추가적인 작업이 요구된다. 또한 기존 방법은 일부 최소 단절군이 생성되지 않는 약점을 갖고 있다. 이에 따라 한국원자력연구소에서는 해석적으로 순환 논리를 푸는 방법을 개발하였으며, 이를 PSA용 코드인 KIRAP 코드에 구축하였다. 이에 따라 기존 방법의 약점을 극복하고 고장 수목간의 순환 논리를 자동으로 풀 수 있게 되었다. 본 논문에서는 개발된 해석적 방법을 설명하며, 또한 이 방법을 실제 PSA에 적용하며 나타난 여러 현상에 대하여 살펴본다.