• Corrosion Science and Technology
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• 제11권6호
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• pp.225-231
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• 2012

Fitness-for-Service is a useful technology to determine replacement timing, next inspection timing or in-service when nuclear power plant's buried pipes are damaged. If is possible for buried pipes to be aged by material loss, cracks and occlusion as operating time goes by. Therefore Fitness-for-Service technology for buried pipe is useful for plant industry to perform replacement and repair. Fitness-for-Service for buried pipe is studied in terms of existing code and standard for Fitness-for-Service and a current developing code case. Fitness-for-Service for buried pipe was performed according to Code Case N-806 developed by ASME (American Society of Mechanical Engineers).

• Nuclear Engineering and Technology
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• 제49권2호
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• pp.327-334
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• 2017

Common-cause failure (CCF) events can significantly impact the availability of safety systems of nuclear power plants. For this reason, the International Common Cause Data Exchange (ICDE) project was initiated by several countries in 1994. Since 1997 it has been operated within the Organisation for Economic Co-operation and Development (OECD)/Nuclear Energy Agency (NEA) framework and has successfully been operated over six consecutive terms (the current term being 2015-2017). The ICDE project allows multiple countries to collaborate and exchange CCF data to enhance the quality of risk analyses, which include CCF modeling. As CCF events are typically rare, most countries do not experience enough CCF events to perform meaningful analyses. Data combined from several countries, however, have yielded sufficient data for more rigorous analyses. The ICDE project has meanwhile published 11 reports on the collection and analysis of CCF events of specific component types (centrifugal pumps, emergency diesel generators, motor operated valves, safety and relief valves, check valves, circuit breakers, level measurement, control rod drive assemblies, and heat exchangers) and two topical reports. This paper presents recent activities and lessons learnt from the data collection and the results of topical analysis on emergency diesel generator CCF impacting entire exposed population.

• Nuclear Engineering and Technology
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• 제46권1호
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• pp.11-26
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• 2014

The analyses carried out within the Seismic Probabilistic Risk Assessments (SPRAs) of Nuclear Power Plants (NPPs) are affected by significant aleatory and epistemic uncertainties. These uncertainties have to be represented and quantified coherently with the data, information and knowledge available, to provide reasonable assurance that related decisions can be taken robustly and with confidence. The amount of data, information and knowledge available for seismic risk assessment is typically limited, so that the analysis must strongly rely on expert judgments. In this paper, a Dempster-Shafer Theory (DST) framework for handling uncertainties in NPP SPRAs is proposed and applied to an example case study. The main contributions of this paper are two: (i) applying the complete DST framework to SPRA models, showing how to build the Dempster-Shafer structures of the uncertainty parameters based on industry generic data, and (ii) embedding Bayesian updating based on plant specific data into the framework. The results of the application to a case study show that the approach is feasible and effective in (i) describing and jointly propagating aleatory and epistemic uncertainties in SPRA models and (ii) providing 'conservative' bounds on the safety quantities of interest (i.e. Core Damage Frequency, CDF) that reflect the (limited) state of knowledge of the experts about the system of interest.

• Nuclear Engineering and Technology
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• 제55권8호
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• pp.2747-2756
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• 2023

Reinforced Concrete (RC) shear walls are one of the civil structures in nuclear power plants to resist lateral loads such as earthquakes and wind loads effectively. Risk-informed and performance-based regulation in the nuclear industry requires considering possible accidents and determining desirable performance on structures. As a result, rather than predicting only the ultimate capacity of structures, the prediction of performances on structures depending on different damage states or various accident scenarios have increasingly needed. This study aims to develop machine-learning models predicting drifts of the RC shear walls according to the damage limit states. The damage limit states are divided into four categories: the onset of cracking, yielding of rebars, crushing of concrete, and structural failure. The data on the drift of shear walls at each damage state are collected from the existing studies, and four regression machine-learning models are used to train the datasets. In addition, the bagging ensemble method is applied to improve the accuracy of the individual machine-learning models. The developed models are to predict the drifts of shear walls consisting of various cross-sections based on designated damage limit states in advance and help to determine the repairing methods according to damage levels to shear walls.

• 한국압력기기공학회 논문집
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• 제18권1호
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• pp.1-10
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• 2022

Currently, there is no technical standard and regulation for seismic analysis of non-nuclear safety piping. Accordingly, ASME Sec.III ND, a standards applied to safety class 3 piping, is applied. However, the technical standard applied for other than seismic analysis is ASME B31, which leads to controversy. In this study, the feasibility of applying ASME B31E was confirmed by reviewing rulescomparing technical standards, and evaluating piping allowable stress margins. The evaluation revealed that applying ASME B31.1 as a technical standard is too conservative compared to ASME Sec.III ND. On the other hand, ASME B31E (issued at the request of the industry) clearly presents the technical standards for seismic analysis of ASME B31 piping, and shows a similar level of conservatism compared to ASME Sec.III ND. It is expected to reduce the controversy over technical standards for seismic analysis of non-nuclear safety piping by applying ASME B31E.

• Nuclear Engineering and Technology
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• 제55권10호
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• pp.3709-3715
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• 2023

It is an important work to measure the dimensions of underwater spent fuel assemblies in the nuclear power industry during the overhaul, to judging whether the spent fuel assemblies can continue to be used. In this paper, a three dimensional reconstruction method for underwater spent fuel assemblies of nuclear reactor based on linear structured light is proposed, and the topography and size measurement was carried out based on the reconstructed 3D model. Multiple linear structured light sensors are used to obtain contour size data, and the shape data of the whole spent fuel assembly can be collected by one-dimensional scanning motion. In this paper, we also presented a corrected model to correct the measurement error introduced by lead-glass and water is corrected. Then, we set up an underwater measurement system for spent fuel assembly based on this method. Finally, an underwater measurement experiment is carried out to verify the 3D reconstruction ability and measurement ability of the system, and the measurement error is less than ±0.05 mm.

• 융합보안논문지
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• 제14권1호
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• pp.11-21
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• 2014

한국은 과거처럼 국제관계의 예속자가 아니라 명실상부한 중견국으로서, 북한의 핵과 재래전 도발위협을 극복하고 동북아지역의 평화를 유지하는 '균형자' 역할을 할 수 있도록 군사력을 건설해야 한다. 군사력 건설을 통해 다양한 안보위협에 대한 억제력 발휘가 가능하다. 군사적 억제력 발휘를 위해 첫 번째로 '선제적 억제'(deterrence by preemptive)와 '응징적 억제'(deterrence by punishment)는 현재와 미래의 위협에 대비하여 '감시정찰체계와 지휘통제체계'(C41SR)를 공통전력으로 공격무기체계를 결합한 '공격체계 축'을 건설함으로써 달성할 수 있다. 두 번째로 '거부적 억제'(deterrence by denial)는 공통전력과 방어무기체계를 결합한 '방어체계 축'을 건설함으로써 달성할 수 있다. 마지막으로 자주적으로 첨단전력을 개발하기 위해서는 기존의 방위산업과 연구개발 역량을 통합하여 '인프라 축'을 구축해야 한다. 우리는 미래 한국군의 군사력을 건설함에 있어서 정부의 균형자 역할에 대한 국가적 비젼, 이에 대한 국민적 합의를 토대로 본고에서 제시한 군사력 건설 모형에 따른 일관성 있는 정책적인 노력과 신념이 반드시 필요하다.

• 방사선산업학회지
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• 제17권3호
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• pp.265-273
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• 2023

To effectively and safely manage the radiation exposure to nuclear power plant (NPP) workers in accidents, major overseas NPP operators such as the United States, Germany, and France have developed and applied realistic 3D model radiation dose assessment software for workers. Continuous research and development have recently been conducted, such as performing NPP accident management using 3D-VR based on As Low As Reasonably Achievable (ALARA) planning tool. In line with this global trend, it is also required to secure technology to manage radiation exposure of workers in Korea efficiently. Therefore, in this paper, it is described the application method and assessment results of radiation exposure scenarios for workers in response to accidents assessment technology, which is one of the fundamental technologies for constructing a realistic platform to be utilized for radiation exposure prediction, diagnosis, management, and training simulations following accidents. First, the post-accident sampling after the Loss of Coolant Accident(LOCA) was selected as the accident and response scenario, and the assessment area related to this work was established. Subsequently, the structures within the assessment area were modeled using MCNP, and the radiation source of the equipment was inputted. Based on this, the radiation dose distribution in the assessment area was assessed. Afterward, considering the three principles of external radiation protection (time, distance, and shielding) detailed work scenarios were developed by varying the number of workers, the presence or absence of a shield, and the location of the shield. The radiation exposure doses received by workers were compared and analyzed for each scenario, and based on the results, the optimal accident response scenario was derived. The results of this study plan to be utilized as a fundamental technology to ensure the safety of workers through simulations targeting various reactor types and accident response scenarios in the future. Furthermore, it is expected to secure the possibility of developing a data-based ALARA decision support system for predicting radiation exposure dose at NPP sites.

• 방사선산업학회지
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• 제9권3호
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• pp.127-130
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• 2015

Since the accident at the Fukushima Daiichi power plant, Prussian blue (PB) has attracted increasing attention as a material for use in decontaminating the environment. We have focused the fundamental mechanism of specific $Cs^+$ adsorption into PB in order to develop high-performance PB-based $Cs^+$ adsorbents. The ability of PB to adsorb Cs varies considerably according to its origin such as what synthesis method was used, and under what conditions the PB was prepared. It has been commonly accepted that the exclusive abilities of PB to adsorb hydrated $Cs^+$ ions are caused by regular lattice spaces surrounded by cyanido-bridged metals. $Cs^+$ ions are trapped by simple physical adsorption in the regular lattice spaces of PB. $Cs^+$ ions are exclusively trapped by chemical adsorption via the hydrophilic lattice defect sites with proton-exchange from the coordination water. Prussian blue are believed to hold great promise for the clean-up of $^{137}Cs$ contaminated water around nuclear facilities and/or after nuclear accidents.

• 방사선산업학회지
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• 제17권1호
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• pp.111-118
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• 2023

The International Commission on Radiological Protection (ICRP) 103 recommends a cost-benefit analysis method as an auxiliary tool for scientific and rational decision-making for the principle of optimization of radiological protection. In order to conduct a cost-benefit analysis, the safety improvement of nuclear power by regulation must be measured and converted into monetary terms. The improvement of nuclear safety can be measured by reducing the radiation exposure dose of the people, and it is necessary to determine the coefficient to convert the radiation exposure dose into money. The monetary coefficient is calculated as the product of the statistical life value (VSL) and the nominal risk coefficient. In order to derive the monetary coefficient, the willingness to pay (WTP) can be estimated using the contingent valuation method (CVM), which quantifies the value of non-market goods by converting them into monetary units. WTP can be estimated based on the random utility model, which is the basic model for bivariate selection type conditional value measurement data. Statistical life value can be calculated using the estimated WTP and reduction in early mortality, and a monetary coefficient can be derived.