• Proceedings of the Korea Information Processing Society Conference
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• 2001.04b
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• pp.1097-1100
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• 2001

소프트웨어 개발자로부터 독립된 소프트웨어 시험자가 수행하는 소프트웨어 시험은 소프트웨어의 안전성 향상을 위해 필요하다. 컴퓨터기반의 디지틀시스템으로 설계되는 일체형원자로 MMIS에 적용하기 위한 소프트웨어 시험 계획을 개발할 필요가 있다. 본 논문은 소프트웨어 시험 계획을 소프트웨어시험 조직 구성, 시험 문서, 시험 절차, 시험 방법을 중심으로 제시한다. 소프트웨어 시험 방법은 원시코드 정적분석과 동적시험을 구분하여 기술한다. 본 논문에서 제시된 소프트웨어 시험 계획은 원자력 규제기관에서 요구하는 소프트웨어 시험 요구사항을 만족한다. 본 논문을 통해 제시된 소프트웨어 시험 계획을 일체형원자로 MMIS 소프트웨어 개발 시 적용하여 소프트웨어 고장율 데이터를 수집할 예정이다.

• The Transactions of the Korean Institute of Power Electronics
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• v.26 no.3
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• pp.222-226
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• 2021

To comply with CE marking requirements, the electromagnetic compatibility (EMC) and low-voltage directive (LVD) tests are conducted on the sub-racks of International Thermonuclear Experimental Reactor (ITER) AC/DC converters and bypass switches. The EMC tests consist of a series of tests, including the electromagnetic interference test, the electromagnetic field immunity test, and the rapid transient burst immunity test. In the LVD test, the electric shock protection test, the xcessive temperature limit and heat resistance of equipment tests, and the fire spread prevention test are performed. This work presents and reviews the European Directive for EMC/LVD and introduces the methods of EMC and LVD tests for the sub-racks of AC/DC converters and bypass switches. It also evaluates the test method and results to meet the European Directive requirements for CE marking. The sub-racks of ITER AC/DC converters and bypass switches successfully pass the EMC and LVD tests.

• Nuclear Engineering and Technology
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• v.56 no.9
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• pp.3626-3643
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• 2024

One of the issues in the development of the sodium-cooled fast reactor (SFR) using transuranic (TRU) metallic fuel is the absence of criticality benchmark experiment that faithfully mocks up the nuclear characteristics of the target design for validation of the reactor core design code and its uncertainty quantification (UQ). This study aims to quantify the criticality uncertainty of a typical TRU burner with metallic fuel by using the standard upper safety limit (USL) estimation framework based on the similarity analysis of existing benchmark experiments but elaborated in two aspects:1) application of two-sided rather than one-sided tolerance interval and 2) inclusion of additional uncertainty to account for fission products and minor actinides not included in the benchmark experiments. To conduct the similarity analysis and evaluate the nuclear-data induced uncertainty, existing, well-verified computing codes were integrated, including the nuclear data sampling code SANDY, the nuclear data processing code NJOY, and the continuous-energy Monte Carlo code McCARD. Finally, using the SFR benchmark database comprising both publicly available and proprietary benchmark experiments, the criticality uncertainty of the TRU core model with metallic fuel was evaluated.

• Journal of the Ergonomics Society of Korea
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• v.32 no.1
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• pp.17-26
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• 2013

Objective: The aim of this study is ergonomic analysis of tele-operation tasks using modified remote handling devices dedicated to the cell of PRIDE(PyRoprocess Integrated inactive DEmonstration facility) in KAERI(Korea Atomic Energy Research Institute). Background: Tele-operation manipulators of the PRIDE are applied to perform the remote handling and management of pyroprocessing facilities. Generally, these kinds of systems are composed of master-slave system and its peripherals installed along a wall or ceiling of the cell, and the manipulators transmit the user's own motion to grippers directly. However, a user convenience and intuitiveness while operating the manipulators have not been fully considered in research fields. Method: This study tries to analyze the ergonomic performance of remote handling manipulators in the developed cell facility. It was included that the analysis of operator's capability for his/her own motion range of upper arm while manipulating the MSM, considerations of its manipulation margin and related tool modifications to improve the remote handling performance. Conclusion: The test results of several remote handling tasks performed in PRIDE are represented, and adequate operation strategies for the tele-operation system of hot-cell type facilities are proposed. Application: The knowledge represented in this study can be utilized to improve a tele-operation system operated in a large-scale hot-cell system.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2004.02a
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• pp.96-105
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• 2004

The decommissioning project of the KRR 1 & 2 was started in January 1997. The actual decommissioning activity was started at the RI production facility and was finished at the end of 2002. The dismantling works of all components including the reactor structure of the KRR-2 was started in January, 2003 and will be carried out for 2 years till the end of 2004. The project schedule is estimated to delay for 4∼5 months beyond the original plan because of delaying on the cutting of thermal column nose and removal of the graphite bricks, but it may be caught up during the removal working of concrete from biological shielding structure. This paper summarizes the general status of the KRR 1 & 2 and decommissioning activities.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2004.02a
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• pp.114-129
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• 2004

The 3D simulations for the Rotary Specimen Rack (RSR), the shielding concret, and the reactor core dismantling processes in the Korea Research Reactor-1&2(KRR-1&2) were carried out in the present work. The four main dismantling items, which are the RSR, reactor core, beam tube, and the thermal column and the shield concrete, were selected among the many components in the KRR-2 by consideration of the activation, worker training, difficulty of the work and so on. On the basis of these, we built 3D CAD models, selected the proper dismantling technologies, and reviewed their dismantling processes. In this study, the 3D simulation results of the shielding concrete, and the reactor core dismantling processes are also presented and discussed.

• Nuclear Engineering and Technology
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• v.39 no.3
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• pp.171-186
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• 2007

Critical issues in the development targets for the future fast reactor(FR) cycle system, including sodium-cooled FR were to ensure safety assurance, efficient utilization of resources, reduction of environmental burden, assurance of nuclear non-proliferation, and economic competitiveness. A promising design concept of sodium-cooled fast reactor JSFR is proposed aiming at fully satisfaction of the development targets for the next generation nuclear energy system. A roadmap toward JSFR commercialization is described, to be followed up in a new framework of the Fast reactor Cycle Technology development(FaCT) Project launched in 2006.

• Proceedings of the Korea Information Processing Society Conference
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• 2010.04a
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• pp.841-843
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• 2010

원자력시설의 해체사업은 국내에서는 최초로 수행된 사업이다. 해체사업 활동에서 발생되는 해체 정보 및 자료의 관리와 사업의 수행을 위해 해체 정보관리 시스템 (DECOMMIS)을 개발하였다. DECOMMIS를 이용하여 해체 작업 활동, 해체 폐기물 관리, 해체 방사선 안전관리 및 품질 관리에 관한 정보/자료를 입력, 처리, 관리하고 필요 시 출력하여 활용하였다.

• Corrosion Science and Technology
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• v.21 no.6
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• pp.466-475
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• 2022

Surface oxides and intergranular (IG) oxidation phenomena in Alloy 600 depending on hydrogen concentration were characterized to obtain clear insight into the primary water stress corrosion cracking (PWSCC) behavior upon exposure to pressurized water reactor primary water. When hydrogen concentration was between 5 and 30 cm³ H₂/kg H₂O, NiFe₂O₄ and NiO type oxides were found on the surface. NiO type oxides were found inside the oxidized grain boundary when hydrogen concentration was 5 cm³ H₂/kg H₂O. However, only NiFe₂O₄ spinel on the surface and Ni enrichment were observed when hydrogen concentration was 30 cm³ H₂/kg H₂O. These results indicate that the oxidation/reduction reaction of Ni in Alloy 600 depending on hydrogen concentration can considerably affect surface oxidation behavior. It appears that the formation of NiO type oxides in a Ni oxidation state and Ni enrichment in a Ni reduction (or metallic) state are common in primary water. It is believed that the above different oxidation/reduction reactions of Ni in Alloy 600 depending on hydrogen concentration can also significantly affect the resistance to PWSCC of Alloy 600.

• Nuclear Engineering and Technology
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• v.38 no.2
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• pp.119-128
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• 2006

We start to develop a predictable technology for thermal-hydraulic performance of the RMWR core using an advanced numerical simulation technology. As a part of this technology development, we are developing the advanced interface tracking method to improve the conservation of volume of fluid. The present paper describes a part of the development of the twophase flow simulation code TPFIT with the advanced interface tracking method. The numerical results applied to large-scale water-vapor two-phase flow in tight lattice rod bundles are shown and compared with experimental results. In the results of numerical simulation, a tendency of the predicted void fraction distribution in horizontal plane agreed with the measured values obtained by the advanced neutron radiography technique including the bridge formation of the liquid at the position of adjacent fuel rods where an interval is the narrowest.