• Journal of Radiation Protection and Research
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• 제23권3호
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• pp.197-210
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• 1998

IAEA는 1996년 방사성물질 안전 운송규정을 개정하였다. 이 규정은 방사성물질의 우송이나 포장에 관하여 우리나라를 비롯한 각국의 운송관련 규정의 기준이 되는 것으로 전반부에서는 IAEA가 1991년에 출간된 국제방사선방호위원회 (ICRP)의 신권고(Publication 60)를 받아들여 개정한 Safety Series No.115(전리방사선에 대한 방호 또는 방사선원의 안전을 위한 기본 안전기준)의 내용 등을 개정의 배경으로 하여 요약하였다. 후반부에서는 이들 개정된 기본 안전 기준들에 기초하여 IAEA의 새로운 운송규정에서 방사선방호의 목적으로 고찰된 요건들에 관한 주요 검토 개정사항을 방사선방어체계, 운송물등의 방사선준위, 방사선방호계획의 규정화, Q 시스템의 개념, 규제면제 등의 측면에서 Safety Series No.6 1985년판과 비교 검토하였다.

• 한국압력기기공학회 논문집
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• 제16권1호
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• pp.117-124
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• 2020

3D printing is a technology that has significantly grown in recent years, particularly in the aerospace, defense, and medical sectors where it offers significant potential cost savings and reduction of the supply chain by allowing parts to be manufactured on-site rather than at a distance supplier. In nuclear industry, 3D printing technology should be applied according to the manufacturing trend change. For the application of 3D printing technology to the nuclear power plant, several problems, including the absence of code & standards of materials, processes and testing & inspection methods etc, should be solved. Preemptively, the improvement of reliability of 3D printing technology, including mechanical properties, structural performance, service performance and aging degradation of 3D printed parts should be supported. These results can be achieved by collaboration of many organizations such as institute, 3D printer manufacturer, metal powder supplier, nuclear part manufacturer, standard developing organization, and nuclear utility.

• 한국건축시공학회:학술대회논문집
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• 한국건축시공학회 2017년도 춘계 학술논문 발표대회
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• pp.200-201
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• 2017

Generally, a lot of reinforcements are used in nuclear power plant concrete structures, and it may cause several potential problems when concrete is poured. Because of the congestion caused by hooked bars, embedded materials, and other reinforcements, it is too difficult to pour concrete into structural member joint area. The purpose of this study is to change ACI 349 Code for using the large-size(57mm) and high-strength(Gr.80) headed deformed bars instead of standard hooked bars in nuclear power plant concrete structures in order to solve the congestion problems.

• 한국해양학회지:바다
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• 제12권1호
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• pp.43-49
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• 2007

원자력발전소 온배수 영향 조사 과정상 문제점을 평가하고 대안으로서 조사 표준 지침을 제시하였다. 관측이나 온배수 확산 모델링 단독으로 이루어진 지금까지의 방법으로는 시간에 따른 온배수에 의한 수온 분포의 변화를 정량적으로 파악하는데 한계가 있으므로, 관측과 온배수 확산 수치모델링의 상호 보완적인 조사를 병행하는 것이 바람직하다. 현장 관측은 원자력발전소 인근 해역의 모든 자연적 열원의 영향을 고려한 열수지 모형의 개념에 근거한 조사가 중요하며, 수치모델링의 결과를 기준 수온분포로서 활용하고자 할 경우에는 수치모델링에 의한 현상 재현이 통계적 유의수준에 도달한 경우에 한정하도록 하였다. 또한, 과거의 순환 및 확산 모델링의 문제점을 개선하기 위한 대안으로서 표준코드의 개발을 제안하였다.

• Nuclear Engineering and Technology
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• 제47권5호
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• pp.559-566
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• 2015

The thermalehydraulic characteristics for the CANadian Deuterium Uranium Flexible (CANFLEX)-burnable poison (BP) fuel channel, which is loaded with a BP at the center ring based on the CANFLEX-RU (recycled uranium) fuel channel, are evaluated and compared with that of standard 37-element and CANFLEX-NU (natural uranium) fuel channels. The distributions of fuel temperature and critical channel power for the CANFLEX-BP fuel channel are calculated using the NUclear Heat Transport CIRcuit Thermohydraulics Analysis Code (NUCIRC) code for various creep rate and burnup. CANFLEX-BP fuel channel has been revealed to have a lower fuel temperature compared with that of a standard 37-element fuel channel, especially for high power channels. The critical channel power of CANFLEX-BP fuel channel has increased by about 10%, relative to that of a standard 37-element fuel channel for 380 channels in a core, and has higher value relative to that of the CANFLEX-NU fuel channel except the channels in the outer core. This study has shown that the use of a BP is feasible to enhance the thermal performance by the axial heat flux distribution, as well as the improvement of the reactor physical safety characteristics, and thus the reactor safety can be improved by the use of BP in a CANDU reactor.

• 대한전기학회:학술대회논문집
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• 대한전기학회 2005년도 추계학술대회 논문집 전기설비전문위원
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• pp.73-75
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• 2005

A number of the power cables arc installed in the containment of the nuclear power plant. According to the IEEE Standard 835, the calculation of the temperature rise shows the operation possibility of power cables in the containment. In this paper, we expect the integrity of the power cables by using the calculation of the temperature rise and the development of the lifetime extension of the cables.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2003년도 춘계학술발표대회 요약집
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• pp.201-201
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• 2003

• 대한전기학회:학술대회논문집
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• 대한전기학회 2002년도 합동 추계학술대회 논문집 정보 및 제어부문
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• pp.537-541
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• 2002

This paper is about the development of the turbine valve actuator efficiency analysis system for the standard power plants. We developed hydraulic power unit and turbine valve actuator controller. We designed control algorithm for turbine valve actuator, implemented and verified it at the industrial plants.

• 대한전기학회:학술대회논문집
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• 대한전기학회 2007년도 추계학술대회 논문집 전력기술부문
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• pp.57-59
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• 2007

When onsite power is supplied from grid due to electrical transient in NPP, the effect of the nuclear plant risk will be increased by the change of grid performance. It is important to analyze the effect for bus voltage of onsite according to grid reliability. Therefore, we analytically accomplish the effect for bus voltage by electrical transient in KSNP.

• Nuclear Engineering and Technology
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• 제30권2호
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• pp.173-179
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• 1998

The corrosion behavior of KOFA cladding which is a standard Zircaloy-4 manufactured by Westinghouse Specialty Metal Plant according to the Siemens/KWU's HCW (Highly Cold Worked) standard Zircaloy-4 specification was analyzed using the oxide measurement data of KOFA fuel irradiated in Kori-2 nuclear power plant. Analysis of the measured KOFA cladding oxidation showed that oxidation of KOFA cladding was lower than the design prediction based upon Siemens/KWU's HCW standard Zircaloy-4 cladding. Although the measured fuel rods have relatively low burnup and oxidation and the amount of the measured data are small, analysis of manufacturing and in-reactor operation conditions of KOFA cladding indicates that the differences in the manufacturing processes and chemical composition of the Siemens/KWU's HCW (Highly Cold Worked) standard Zircaloy-4 and KOFA cladding may have somewhat contributed to lower corrosion of KOFA cladding than the expected.