• Nuclear Engineering and Technology
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• v.53 no.8
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• pp.2682-2695
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• 2021

This paper aims to evaluate the structural dynamic responses and damage/failure of the nuclear fuel reprocessing plant under the free drop impact of spent fuel cask (SFC) and fuel assembly (FA) during the on-site transportation. At the present Part I of this paper, the large-scale SFC model free drop test and the corresponding numerical simulations are performed. Firstly, a composite target which is composed of the protective structure, i.e., a thin RC plate (representing the inverted U-shaped slab in the loading shaft) and/or an autoclaved aerated concrete (AAC) blocks sacrificial layer, as well as a thick RC plate (representing the bottom slab in the loading shaft) is designed and fabricated. Then, based on the large dropping tower, the free drop test of large-scale SFC model with the mass of 3 t is carried out from the height of 7 m-11 m. It indicates that the bottom slab in the loading shaft could not resist the free drop impact of SFC. The composite protective structure can effectively reduce the damage and vibrations of the bottom slab, and the inverted U-shaped slab could relieve the damage of the AAC blocks layer dramatically. Furthermore, based on the finite element (FE) program LS-DYNA, the corresponding refined numerical simulations are performed. By comparing the experimental and numerical damage and vibration accelerations of the composite structures, the present adopted numerical algorithms, constitutive models and parameters are validated, which will be applied in the further assessment of drop impact effects of full-scale SFC and FA on prototype nuclear fuel reprocessing plant in the next Part II of this paper.

• Proceedings of the Korean Nuclear Society Conference
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• 1995.10a
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• pp.462-468
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• 1995

SB-LOCA, RCP Failure 등의 특정사고와 운전중의 과도현상의 발생시, 형성되는 자연대류의 PWR의 열수력학적 거동은 로심 잔열제거의 가능/불가능과 직결되므로 매우 중요한 PWR 안전성평가자료이다. 이러한 자료를 획득하기 위해, EG&G의 감독하에 수행된 Semiscale Experiment Program의 일부인 Semiscale Mod-2A실험들 중에서 S-NC-3과 5-NC-4를 REALAP5/MOD3로 모사하였다. 모사결과는 실험결과와 REALAP5/MODl을 이용한 모사결과와 비교 검토하였다.

• 제어로봇시스템학회:학술대회논문집
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• 2000.10a
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• pp.379-379
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• 2000

The robot systems used in nuclear power plants need to be both reliable and safe. As a part of the "Validation of nuclear safety-grade equipment" project, we established reliability analysis program and performed a number of analysis using conventional reliability analysis techniques. This paper describes the procedures, techniques, and results of the analysis utilized in our project. In addition, the paper includes current status of reliability analysis techniques and the summary of foreign case studiesse studies

• Proceedings of the Earthquake Engineering Society of Korea Conference
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• 1997.10a
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• pp.82-89
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• 1997

Due to aging of the imported seismic monitoring system of Uljin 1＆2 units it is difficult for this system to provide enough functions needed for the security of seismic safety and the evaluation of the earthquake data from the seismic instrumentation. For this reason, it is necessary to replace the seismic monitoring system of Uljin 1＆2 units with an upgraded system with corresponding softwars. With operation of this system which incorporates the man-machine interface technology, the operators in nuclear power plant can rapidly and correctly determine the exceedance of Operating Basis Earthquake.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.10a
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• pp.363-368
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• 1997

PWR 원전에서는 계획예방 정지운전시 효과적인 정지수화학 제어에 의해 일차계통 방사능 준위를 감소시키고 작업종사자의 피폭을 최소화하기 위해 정지운전 자료에 대한 보다 정확한 해석이 필요하다. 본 연구에서는 국내 PWR 원전 주기(A호기의 I 및 II주기와 B호기의 II주기)별 정지수화학 특성을 SCALP(Shutdown chemistry CALculation Program)프로그램으로 계산하고 정지운전 기간중 일차냉각재계통에서 제거되는 방사능량에 영향을 미치는 정지수화학 특성을 주요 인자별로 평가하였다.

• Nuclear Engineering and Technology
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• v.26 no.2
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• pp.324-336
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• 1994

경수로형 원전에서 발생되는 작업종사자에 대한 피폭선량의 약90%는 1차계통내 재질성분으로부터 방사화되는 코발트 핵종에 기인한다. 원전 선진국에서는 이러한 코발트 생성원을 근원적으로 제거하여 1차계통내 방사선량을 획기적으로 감소시킬 목적으로 Co reduction program을 중점적으로 추진하여 왔으며, 이중 계통내 코발트 주생성원인 각종 밸브재질을 저 Co 또는 Co-free 합금으로 대체시키는 기술이 이미 상당한 수준에 이르렀다. 국내의 건설예정 원전에 있어서도 이러한 기술개발의 적용에 대한 검토가 요구되고 있는 점을 미루어 볼 때, 머지 않은 장래에 국내 원전에 대한 이 기술의 적용이 불가피할 것으로 보여진다. 본 기고문에서는 원전 1차계통내 밸브의 코발트 합금재질 대체기술과 관련된 내용을 중점적으로 소개, 검토하고자 한다.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05a
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• pp.499-504
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• 1996

본 논문은 저온정지에서 2%까지 원전 자동기동시스템에 필요한 발전소 데이터를 시험검증설비로부터 얻기 위해 공유메모리와 TCP/IP를 사용하여 통신프로그램을 개발하였다. 자동기동시스템은 foxboro에서 제공하는 API(Application Program Interface)를 이용하여 데이터베이스에서 제어기에 필요한 데이터를 공유메모리에 올려놓고, 통신프로그램이 읽고 쓸 수 있도록 했으며, 시험검증설비에서는 HP Workstation에서 사용되는 내부 프로세스 통신방법을 이용하여 시험검증설비에서 나온 데이터를 공유메모리에 넣을 수 있도록 설계하였다.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.11a
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• pp.211-216
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• 1996

Since the middle of the eighties of this century a research program both for heating reactor and investigation of heating reactor thermal-hydraulics has been carried out in Institute of Nuclear Energy Technology(INET) of Tsinghua university in China. This kind of heating reactor is a light water cooled and integrated natural circulation reactor with low system pressure and low quality at the exit of core. Because of relatively long riser and low system pressure. a little change of the quality at the exit of the core will result in a relatively large variation of void fraction in the riser. Two full scale test loops. HRTL-5 and HRTL-200 simulating the HR-5 and HR-200 heating reactors in geometry and operation parameters respectively, and some test results from the HRTL-200 test facility are shown in this paper. The range of studied system pressure is from 1.0MPa to 4.0MPa, the largest heat flux is about 50 W/cm2, and the quality at the exit of test section is less than 5%.

• Journal of Energy Engineering
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• v.26 no.2
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• pp.12-22
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• 2017

Korea Hydro & Nuclear Power Co. (KHNP) has implemented engineering programs for operating nuclear power plants. Engineering programs are maintenance rule (MR), functional importance determination (FID), single point vulnerability (SPV) and functional equipment group (FEG). Recently, KHNP has developed engineering programs for APR1400 feedwater supplying system to establish the advanced engineering system and will verify the suitability of engineering programs through implementing in new nuclear power plant. Consequently, it is expected that the reliability of APR1400 feedwater supplying system will be improved by implementing engineering programs.

• Nuclear Engineering and Technology
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• v.41 no.4
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• pp.545-548
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• 2009

Korean utility has utilized a Periodic Safety Review (PSR) that assesses the cumulative effects of plant aging, modifications, operating experience, technical developments, and site characteristics since 2000. In particular, the assessment and management of plant aging is one of the major areas in PSR. It includes identification of critical Systems, Structures, and Components (SSCs) for aging, assessment of aging effects, and implementation of aging management programs. Since the PSR system was introduced based on the atomic energy acts and related laws, PSRs of eight sets for 12 Nuclear Power Plants (NPPs) that have been operating more than 10 years have been completed. PSRs of two sets for 4 NPPs are currently being carried out. The utility has confirmed that domestic NPPs have been operated safely through these PSRs and have implemented the follow-up corrective activities to increase the nuclear safety. In this paper, the status of PSR implementation is discussed and improvement programs to conduct PSR follow-up corrective activities efficiently for NPPs are suggested based on experiences with aging assessments.