• Nuclear Engineering and Technology
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• 제52권1호
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• pp.192-205
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• 2020

Seismic design practices and seismic response analyses of civil structures and nuclear power plants (NPPs) have conventionally used the peak ground acceleration (PGA) or spectral acceleration (S_a) as an intensity measure (IM) of an earthquake. However, there are many other earthquake IMs that were proposed by various researchers. The aim of this study is to investigate the correlation between seismic responses of NPP components and 23 earthquake IMs and identify the best IMs for correlating with damage of NPP structures. Particularly, low- and high-frequency ground motion records are separately accounted in correlation analyses. An advanced power reactor NPP in Korea, APR1400, is selected for numerical analyses where containment and auxiliary buildings are modeled using SAP2000. Floor displacements and accelerations are monitored for the non- and base-isolated NPP structures while shear deformations of the base isolator are additionally monitored for the base-isolated NPP. A series of Pearson's correlation coefficients are calculated to recognize the correlation between each of the 23 earthquake IMs and responses of NPP structures. The numerical results demonstrate that there is a significant difference in the correlation between earthquake IMs and seismic responses of non-isolated NPP structures considering low- and high-frequency ground motion groups. Meanwhile, a trivial discrepancy of the correlation is observed in the case of the base-isolated NPP subjected to the two groups of ground motions. Moreover, a selection of PGA or S_a for seismic response analyses of NPP structures in the high-frequency seismic regions may not be the best option. Additionally, a set of fragility curves are thereafter developed for the base-isolated NPP based on the shear deformation of lead rubber bearing (LRB) with respect to the strongly correlated IMs. The results reveal that the probability of damage to the structure is higher for low-frequency earthquakes compared with that of high-frequency ground motions.

• 방사성폐기물학회지
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• 제16권3호
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• pp.331-337
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• 2018

원전 해체는 일반적으로 5단계로 준비, 제염, 절단 및 철거, 폐기물 처리, 환경 복원으로 진행된다. 효율적인 원전 해체를 위해서는 작업자의 안전, 비용 대비 효과, 폐기물 최소화, 재사용 가능성 등이 고려되어야 한다. 또한, 작업자의 안전 및 측정기술이 확보되어야 원전 해체 작업의 최적 효율을 낼 수 있으며 이를 위해서는 계통 및 기기의 정확한 측정 기술이 필요하다. 원전 해체 시 현장에서 사용할 수 있는 대표적인 In-Situ 방법으로는 CZT, Gamma Camera, ISOCS 등이 있다. 본 연구에서는 대표 시료 채취 없이 원전 해체 시 현장에서 적용될 수 있는 ISOCS를 이용하여 S/G Water Chamber 지점에 대하여 측정을 수행하였다. 측정 방법은 ISOCS의 HPGe 검출기를 증기 발생기 수실 하부 중앙을 향해 위치하였으며, 이때 검출기는 주변 방사선장 감소를 위해 납 차폐체를 장착하였다. 차폐체 두께는 5 cm인 원통형 납 차폐체를 장착하였으며, 검출기 전면에는 30도 콜리메이터를 장착하여 측정을 수행하였다. 측정값에 검증을 위해 실제 측정 방법과 동일하게 Microshield를 이용하여 측정한 값과 GEANT4 코드를 이용하여 모델링 하였다. 비교 결과 $1.0{\times}10^1{\sim}1.0{\times}10^2Bq$ 정도 차이를 보였으며, 이는 측정 시 주변 방사선의 영향, 모델링의 정밀도 등으로 오차를 줄일 수 있을 것으로 보인다. 본 논문의 연구 결과를 바탕으로 측정값의 정확도 및 신뢰도를 분석하고 향후 해체 작업 시 직접 측정 방법의 적용성에 대한 신뢰도를 높이고자 한다.

• 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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• 제53권2호
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• pp.466-473
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• 2021

Fire protection is one of important issues to ensure safety and reduce risks of nuclear power plants (NPPs). While robust programs to shut down commercial reactors in any fires have been successfully maintained, the concept and associated regulatory requirements are constantly changing or strengthening by lessons learned from operating experiences and information all over the world. As part of this context, it is necessary not only to establish specific fire hazard assessment methods reflecting the characteristics of research reactors and educational reactors but also to make decisions based on advancement encompassing numerical analyses and experiments. The objectives of this study are to address fire simulation in the control room of an educational reactor and to discuss integrity of digital console in charge of main operation as well as analysis results through comparison. Three electrical fire scenarios were postulated and twenty-four thermal analyses were carried out taking into account two turbulence models, two cable materials and two ventilation conditions. Twelve supplementary thermal analyses and six subsequent structural analyses were also conducted for further examination on the temperature and heat flux of cable and von Mises stress of digital console, respectively. As consequences, effects of each parameter were quantified in detail and future applicability was briefly discussed. On the whole, higher profiles were obtained when Deardorff turbulence model was employed or polyvinyl chloride material and larger ventilation condition were considered. All the maximum values considered in this study met the allowable criteria so that safety action seems available by sustained integrity of the cable linked to digital console within operators' reaction time of 300 s.

• 방사성폐기물학회지
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• 제18권4호
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• pp.481-496
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• 2020

After nuclear power plants are permanently shut down and decommissioned, the remaining irradiated metal components such as stainless steel, carbon steel, and Inconel can be used as neutron absorber. This study investigates the possibility of reusing these metal components as neutron absorber materials, that is burnable poison. The absorption cross section of the irradiated metals did not lose their chemical properties and performance even if they were irradiated over 40-50 years in the NPPs. To examine the absorption capability of the waste metals, the lattice calculations of WH 17×17 fuel assembly were analyzed. From the results, Inconel-718 significantly hold-down fuel assembly excess reactivity compared to stainless steel 304 and carbon steel because Inconel-718 contains a small amount of boron nuclide. From the results, a 20wt% impurity of boron in irradiated Inconel-718 enhances the excess reactivity suppression. The application of irradiated Inconel-718 as a burnable absorber for SMR core was investigated. The irradiated Inconel-718 impurity with 20wt% of boron content can maintain and suppress the whole core reactivity. We emphasize that the irradiated metal components can be used as burnable absorber materials to control the reactivity of commercial reactor power and small modular reactors.

• Nuclear Engineering and Technology
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• 제55권10호
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• pp.3543-3548
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• 2023

Shutdown chemistry evolution is performed in nuclear power plants at each refueling outage (RFO) to establish safe conditions to open system and minimize inventory of corrosion products in the reactor coolant system (RCS). After hydrogen peroxide is added to RCS during shutdown chemistry evolution, corrosion products are released and are removed by filters and ion exchange resins in the chemical volume control system (CVCS). Shutdown chemistry evolution including RCS clean-up time to remove released corrosion products impacts the critical path schedule during RFOs. The estimation of clean-up time prior to RFO can provide more reliable actions for RCS clean-up operations and transients to operators during shutdown chemistry. Electric Power Research Institute (EPRI) shutdown calculator (SDC) enables to provide clean-up time by Co-58 peak activity through operational data from nuclear power plants (NPPs). In this study, we have investigated the results of EPRI SDC by shutdown chemistry data of Co-58 activity using NPP data from previous cycles and modeled the estimated clean-up time by EPRI SDC using average Co-58 activity of the NPP. We selected two RFO data from the NPP to evaluate EPRI SDC results using the purification time to reach to 1.3 mCi/cc of Co-58 after hydrogen peroxide addition. Comparing two RFO data, the similar purification time between actual and computed data by EPRI SDC, 0.92 and 1.74 h respectively, was observed with the deviation of 3.7-7.2%. As the modeling the estimated clean-up time, we calculated average Co-58 peak concentration for normal cycles after cycle 10 and applied two-sigma (2σ, 95.4%) for predicted Co-58 peak concentration as upper and lower values compared to the average data. For the verification of modeling, shutdown chemistry data for RFO 17 was used. Predicted RCS clean-up time with lower and upper values was between 21.05 and 27.58 h, and clean-up time for RFO 17 was 24.75 h, within the predicted time band. Therefore, our calculated modeling band was validated. This approach can be identified that the advantage of the modeling for clean-up time with SDC is that the primary prediction of shutdown chemistry plans can be performed more reliably during shutdown chemistry. This research can contribute to improving the efficiency and safety of shutdown chemistry evolution in nuclear power plants.

• Nuclear Engineering and Technology
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• 제55권8호
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• pp.3054-3070
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• 2023

Nuclear Power Plants (NPP) are widely utilized around the globe from different base forms as it is one of the most dependable renewable resources that technological advancements have offered. However, different perceptions of the usage of NPPs emerged from different generations. The purpose of this study was to investigate the acceptance of nuclear energy as an alternative source of energy among Generation Z in the Philippines by utilizing an extended Theory of Planned Behavior (TPB) approach. An online questionnaire which consisted of 31 items was distributed using a purposive sampling approach and 450 respondents of Generation Z voluntarily answered. Structural Equation Modeling (SEM) showed that the knowledge regarding NPP had significant effects on risk perception and benefit perception which subsequently led to subjective norms. In addition, perceived behavioral control and subjective norms had significant effects on behavioral intention which led to nuclear acceptance. Interestingly, the respondents perceived the benefit of NPP as slightly higher than the perceived risk. With these, it was clear that the commissioning Nuclear Power Plant must consider as an alternative source of electric energy in the Philippines. Moreover, this study is one of the first studies that investigated the acceptance of NPP among Generation Z. Lastly, the model could be a basis to strengthen the acceptance strategy of opening NPP among Generation Z, particularly in developing countries.

• 대한전기학회:학술대회논문집
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• 대한전기학회 2006년 학술대회 논문집 정보 및 제어부문
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• pp.600-602
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• 2006

Some abnormal signals diagnostics and analysis through an important equivalent circuits modeling for passive elements under severe accident conditions have been performed. Unlike the design basis accidents, there are inherently some uncertainties in the instrumentation capabilities under the accident conditions. So, the circuit simulation analysis and diagnosis methods are used to assess instruments in detail when they give apparently abnormal readings as an accident alternative method. The simulations can be useful to investigate what the signal and circuit characteristics would be when similar to a variety of symptoms that can result from the environmental conditions such as high temperature, humidity, and pressure condition. In this paper, a new simulator through an analysis of the important equivalent circuits modeling under temperature accident conditions has been designed, the designed simulator is composed of the LabVIEW code as a main tool and the out-put file of the Multi-SIM code as an engine tool is exported to in-put file of the LabVIEW code. The procedure for the simulator design was divided into two design steps, of which the first step was the diagnosis method, the second step was the circuit simulator for the signal processing tool. It has three main functions which are a signal processing tool, an accident management tool, and an additional guide from the initial screen. This simulator should be possible that it could be applied a output signal analysis to some transient signal by variation of the resistance passive elements in the R-L-C equivalent circuit modeling under various degraded conditions in NPPs.

• 대한인간공학회지
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• 제32권1호
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• pp.59-68
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• 2013

Objective: The purpose of this paper is to introduce a task complexity model combining task design aspects and complexity dimensions and to explain an approach to identifying and organizing task complexity factors based on the model. Background: Task complexity is a critical concept in describing and predicting human performance in complex systems such as nuclear power plants(NPPs). In order to understand the nature of task complexity, task complexity factors need to be identified and organized in a systematic manner. Although several methods have been suggested for identifying and organizing task complexity factors, it is rare to find an analytical approach based on a theoretically sound model. Method: This study regarded a task as a system to be designed. Three levels of design ion, which are functional, behavioral, and structural level of a task, characterize the design aspects of a task. The behavioral aspect is further classified into five cognitive processing activity types(information collection, information analysis, decision and action selection, action implementation, and action feedback). The complexity dimensions describe a task complexity from different perspectives that are size, variety, and order/organization. Combining the design aspects and complexity dimensions of a task, we developed a model from which meaningful task complexity factors can be identified and organized in an analytic way. Results: A model consisting of two facets, each of which is respectively concerned with design aspects and complexity dimensions, were proposed. Additionally, twenty-one task complexity factors were identified and organized based on the model. Conclusion: The model and approach introduced in this paper can be effectively used for examining human performance and human-system interface design issues in NPPs. Application: The model and approach introduced in this paper could be used for several human factors problems, including task allocation and design of information aiding, in NPPs and extended to other types of complex systems such as air traffic control systems as well.

• Journal of Radiation Protection and Research
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• 제36권3호
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• pp.154-159
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• 2011

국내 원전에서는 고 방사선량율 또는 고피폭 예상 방사선작업에 종사자의 가슴과 등에 두 개의 개인선량계를 패용하여 피폭방사선량을 평가하고 있다. 이러한 Two-Dosimeter Algorithm (TDA)으로 현장시험과 심층검토를 통해 NCRP(55:50) TDA를 최적 알고리즘으로 최종적으로 선정하였고, 2006년 이후 원전 종사자의 피폭방사선량 평가 실무에 적용 중에 있다. 한편, 2007년 국제방사선방호위원회(ICRP)는 간행물 ICRP 103을 통해 방사선가중계수 및 조직가중계수와 기준 인체모형팬텀(Reference phantom) 등을 일부 변경한 유효선량 평가방법을 제시하였다. 이에 따라 본 논문에서는 국내원전에서 적용되고 있는 NCRP (55:50) TDA에 대해 ICRP 103 방사선방호 체계 하에서의 계속 적용 타당성을 분석하였다. 그 결과, NCRP (55:50) TDA를 계속 사용하더라도 ICRP 103의 유효선량을 신뢰성 있게 평가할 수 있는 것으로 판단되었다.