• Nuclear Engineering and Technology
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• v.52 no.6
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• pp.1180-1187
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• 2020

A team of operators is required for nuclear power plant operation, and communication between the operators is an important aspect of the team's ability to successfully carry out tasks. It has been difficult to evaluate the quality of this communication though, and as the relationship between communication quality and team performance has yet to be clarified, it has not been applied to most human reliability analysis (HRA) methodologies. This study investigates the relationship between the quality of communication and team performance using data from a full-scope training simulator of a digital main control room (MCR). Two important characteristics of communication were considered to determine quality: each operator's ability to self-confirm the status of a given task in a digital MCR, and the type of communication, as divided into 1-way, 2-way, and 3-way between operators. To measure team performance, the concept of an unsafe act was employed, which is defined as a human error that has the potential to negatively affect plant safety. Analysis results showed that the communication quality and team performance were related to each other. With this more clearly defined relationship, the results of this study can be applied to related performance shaping factors to improve HRA.

• Nuclear industry
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• v.37 no.3
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• pp.77-84
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• 2017

1946년에 캐나다의 원자력에너지관리법(Atomic Energy Control Act)이 제정된 지 70년이 지났다. 후쿠시마 원전 사고 이후에 캐나다에서 시행된 안전조치의 마무리 작업과 주요 원전의 가동 연한 연장에 관한 업무 등 캐나다 원자력안전위원회가 최근 요약해서 발표한 각 부문 업무 현황의 내용을 알아본다.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.15 no.4
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• pp.301-312
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• 2017

The concept of deep borehole disposal (DBD) for high-level nuclear wastes has been around for about 40 years. Now, the Department of Energy (DOE) in the United States (U.S.) is re-examining this concept through recent studies at Sandia National Laboratory and a field test. With DBD, nuclear waste will be emplaced in boreholes at depths of 3 to 5 km in crystalline basement rocks. Thinking is that these settings will provide nearly intact rock and fluid density stratification, which together should act as a robust geologic barrier, requiring only minimal performance from the engineered components. The Nuclear Waste Technical Review Board (NWTRB) has raised concerns that the deep subsurface is more complicated, leading to science, engineering, and safety issues. However, given time and resources, DBD will evolve substantially in the ability to drill deep holes and make measurements there. A leap forward in technology for drilling could lead to other exciting geological applications. Possible innovations might include deep robotic mining, deep energy production, or crustal sequestration of $CO_2$, and new ideas for nuclear waste disposal. Novel technologies could be explored by Korean geologists through simple proof-of-concept experiments and technology demonstrations.

• Nuclear Engineering and Technology
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• v.54 no.11
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• pp.4170-4180
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• 2022

The task complexity (TACOM) measure was previously developed to quantify the complexity of proceduralized tasks conducted by nuclear power plant operators. Following the development of the TACOM measure, its appropriateness has been validated by investigating the relationship between TACOM scores and three kinds of human performance data, namely response times, human error probabilities, and subjective workload scores. However, the information reflected in quantified TACOM scores is still insufficient to determine the levels of complexity of proceduralized tasks for human reliability analysis (HRA) applications. In this regard, the objective of this study is to suggest criteria for determining the levels of task complexity based on logistic regression between human error occurrences in digitalized main control rooms and TACOM scores. Analysis results confirmed that the likelihood of human error occurrence according to the TACOM score is secured. This result strongly implies that the TACOM measure can be used to identify the levels of task complexity, which could be applicable to various research domains including HRA.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.21 no.1
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• pp.1-8
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• 2023

The major concern in the deep geological disposal of spent nuclear fuels include sulfide-induced corrosion and stress corrosion cracking of copper canisters. Sulfur diffusion into copper canisters may induce copper embrittlement by causing Cu₂S particle formation along grain boundaries; these sulfide particles can act as crack initiation sites and eventually cause embrittlement. To prevent the formation of Cu₂S along grain boundaries and sulfur-induced copper embrittlement, copper alloys are designed in this study. Alloying elements that can act as chemical anchors to suppress sulfur diffusion and the formation of Cu₂S along grain boundaries are investigated based on the understanding of the microscopic mechanism of sulfur diffusion and Cu₂S precipitation along grain boundaries. Copper alloy ingots are experimentally manufactured to validate the alloying elements. Microstructural analysis using scanning electron microscopy with energy dispersive spectroscopy demonstrates that Cu₂S particles are not formed at grain boundaries but randomly distributed within grains in all the vacuum arc-melted Cu alloys (Cu-Si, Cu-Ag, and Cu-Zr). Further studies will be conducted to evaluate the mechanical and corrosion properties of the developed Cu alloys.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.6 no.4
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• pp.357-368
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• 2008

On July 31, 2008, the Government issued the construction and operation permit for the first low and intermediate level radioactive waste disposal facility in the Republic of Korea. In this paper, the fundamental regulatory framework, regulatory requirements and technical standards of the disposal facility are introduced, and the phased review process adopted for evaluation of the safety of the facility is briefly described. The Atomic Energy Act sets forth a stepwise regulatory framework for the whole life-cycle of the disposal facility such as siting, design, construction, operation, closure and institutional control. More detailed regulatory requirements and technical standards are stipulated in the subsequent regulations of the Atomic Energy Act and a series of Notices issued by the Ministry of Eduction, Science and Technology. The Korea Institute of Nuclear Safety, as entrusted by the Ministry under the Atomic Energy Act, conducted safety review on the disposal facility, and evaluated the compliance with relevant criteria in all technical elements(i.e. siting and structural safety, radiological environmental impact, operational safety, systems and components, quality assurance, and total systematic performance assessment, etc.). The overall safety review process can be phased into inception phase, initial review phase, main review phase and completion phase. The review results were reported to and deliberated by the five Sub-committees of the Special Committee on Nuclear Safety, and then reported to the Ministry. The Ministry issued the construction and operation permit of the disposal facility through the deliberation of the review results by the Nuclear Safety Commission. Hereafter, the safety of the repository will be reassured by a series of subsequent regulatory inspections and reviews under the Atomic Energy Act. In addition, the licensee's continuous implementation of the "Safety Promotion Plan" may also enhance the long-term safety of the repository and contribute to build-up the confidence of the safety case.

• Journal of the Korean Society of Safety
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• v.26 no.4
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• pp.87-95
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• 2011

The urgent VAI(Vital Area Identification) method development is required since 'The Act of Physical Protection and Radiological Emergency' that is established in 2003 requires an evaluation of physical threats in nuclear facilities and an establishment of physical protection in Korea. The KAERI(Korea Atomic Energy Research Institute) has developed the VAI methodology and VAI software called as VIPEX(Vital area Identification Package EXpert) for identifying the vital areas. This study is to demonstrate the applicability of KAERI's VAI methodology to a hypothetical facility, and to identify the importance of information of cable and piping runs when identifying the vital areas. It is necessarily needed to consider cable and piping runs to determine the accurate and realistic TEPS(Top Event Prevention Set). If the information of cable and piping runs of a nuclear power plant is not considered when determining the TEPSs, it is absolutely impossible to acquire the complete TEPSs, and the results could be distorted by missing it. The VIPEX and FTREX(Fault Tree Reliability Evaluation eXpert) properly calculate MCSs and TEPSs using the fault tree model, and provide the most cost-effective method to save the VAI and physical protection costs.

• Nuclear Engineering and Technology
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• v.53 no.8
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• pp.2582-2590
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• 2021

The high-energy milling is one of the most extended techniques to produce Oxide dispersion strengthened (ODS) powder steels for nuclear applications. The consequences of the high energy mill process on the final powders can be measured by means of deformation level, size, morphology and alloying degree. In this work, an ODS ferritic steel, Fe-14Cr-5Al-3W-0.4Ti-0.25Y₂O₃-0.6Zr, was fabricated using two different mechanical alloying (MA) conditions (M_std and M_act) and subsequently consolidated by Spark Plasma Sintering (SPS). Milling conditions were set to evidence the effectivity of milling by changing the revolutions per minute (rpm) and dwell milling time. Differences on the particle size distribution as well as on the stored plastic deformation were observed, determining the consolidation ability of the material and the achieved microstructure. Since recrystallization depends on the plastic deformation degree, the composition of each particle and the promoted oxide dispersion, a dual grain size distribution was attained after SPS consolidation. M_act showed the highest areas of ultrafine regions when the material is consolidated at 1100 ℃. Microhardness and small punch tests were used to evaluate the material under room temperature and up to 500 ℃. The produced materials have attained remarkable mechanical properties under high temperature conditions.

• Transactions of the Korean Society for Noise and Vibration Engineering
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• v.27 no.2
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• pp.148-154
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• 2017

This paper discussed design for resonance avoidance of sensor plates of loose-parts monitoring systems (LPMS) in nuclear power plants (NPP). An LPMS monitors impact of loose parts in primary loop of NPP by using accelerometers, which is mounted on sensor plates. Resonance of the plates may cause false alarms at frequencies over 10 kHz, which can be misunderstood as impact signals of loose parts with small mass and cause unnecessary response of NPP operators. Modal analysis was carried out for the existing sensor plate and design parameters affecting natural frequencies were chosen. Frequency response functions of plates were analyzed by changing the parameters and the optimized plate design for avoiding resonance was determined. Experiments was carried out for the plate specimen with improved design and verified the proposed approach and design.

• Journal of the Korean Society of Safety
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• v.24 no.5
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• pp.63-68
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• 2009

The urgent VAI method development is required since "The Act of Physical Protection and Radiological Emergency that is established in 2003" requires an evaluation of physical threats in nuclear facilities and an establishment of physical protection in Korea. The VAI methodology is developed to (1) make a sabotage model by reusing existing fire/flooding/pipe break PSA models, (2) calculate MCSs and TEPSs, (3) select the most cost-effective TEPS among many TEPSs, (4) determine the compartments in a selected TEPS as vital areas, and (5) provide protection measures to the vital areas. The developed VAI methodology contains four steps, (1) collecting the internal level 1 PSA model and information, (2) developing the fire/flood/pipe rupture model based on level 1 PSA model, (3) integrating the fire/flood/pipe rupture model into the sabotage model by JSTAR, and (4) calculating MCSs and TEPS. The VAT process is performed through the VIPEX that was developed in KAERI. This methodology serves as a guide to develop a sabotage model by using existing internal and external PSA models. When this methodology is used to identify the vital areas, it provides the most cost-effective method to save the VAI and physical protection costs.