• Proceedings of the Korean Nuclear Society Conference
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• 2002.10a
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• pp.210.1-210
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• 2002

• Proceedings of the Korean Nuclear Society Conference
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• 2003.10a
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• pp.447-447
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• 2003

• Proceedings of the Korean Nuclear Society Conference
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• 2003.10a
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• pp.180-180
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• 2003

• Proceedings of the Korean Nuclear Society Conference
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• 1999.10a
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• pp.201.2-201
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• 1999

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2015.10a
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• pp.527-528
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• 2015

• Proceedings of the Korea Concrete Institute Conference
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• 1993.04a
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• pp.219-224
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• 1993

본 연구는 철근콘크리트 격납구조물에서 가상의 냉각제 유출사고에 의한 온도하중과 압력에 따른 거동을 알아보기 위한 비선형 해석을 수행하였다. 시간에 따른 온도하중을 결정하기 위하여 과도온도해석을 통해 격납구조물 단면내의 온도분포를 구하였다. 구조물은 기하학적 비선형성과 재료비선형성을 고려한 판 및 쉘요소로 이상화되며, 쉘요소는 두께방향에 따라 변하는 응력을 고려하기 위해 몇 개의 층으로 이루어진 모델을 사용하였다. 열응력은 인접한 두시간 단계에서의 온도차를 하나의 하중증가로 고려하여 초기변형 문제로 변환하여 결정되었다. 본연구에서의 수치실험에 의하여 과도온도해석에 근거한 비선형온도경사를 고려할 때의 변위가 고려하지 않을 때의 변위에 비해 크게 나타남을 알 수 있다.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.05a
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• pp.617-622
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• 1997

격납건물내 소격실에서의 수소혼합 정도를 파악하고 격실내 균일한 혼합을 좌우하는 인자의 영향을 분석하기 위하여 소규모 혼합실험을 수행했다. 본 연구에서는 해석적으로 수립된 3차원 혼합 모델의 검증을 위하여 3차원 모사가 가능하도록 실험 장치를 구성하였다. 격납용기 내에서 수소 생성의 주원인이 되는 노심으로부터의 수소거동을 분석하기 위한 기초 실험(실험 A)과 안전주입 탱크 격실에서의 수소거동을 분석하기 위해 원형 혼합 chamber를 구상했다. 기초실험 A에서는 혼합 chamber내 축 방향으로 대칭적인 오리피스형 장애물을 설치하고 실험했고 안전주입 탱크 격실을 모사한 실험 B는 영광 3&4호기를 바탕으로 축소시켜 안전주입탱크 격실내 존재하는 두충과 안전 주입 탱크 사이의 틀을 통한 혼합체의 거동을 분석했다. 실험결과 오리피스형 장애물을 설치한 기초실험에서는 원형 띠모양의 장애물이 혼합체의 거동에 큰 영향을 주지 않는 것이 관측됐지만 안전주입탱크격실 실험에서는 격실내 장애물로 존재하는 두충이 혼합체의 거동에 큰 영향을 주는 것이 관측됐다.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.3 no.1
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• pp.23-30
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• 2005

From measured results of the neutron fields at some principal places within the containment building in a CE type nuclear power plant in operation, the radiation exposure of a worker to the neutron at there was evaluated and the equivalent dose reflecting new recommendation (ICRP 60) was compared with that doing the old one (ICRP 26). The measured neutron field was also compared with calibration neutron field. From the analysis, the following conclusion was obtained: the average neutron radiation weighting factor according to new recommendation is 2.41 to 2.71 times higher than the old one. The average neutorn radiation weighting factor at the measured place was similar to that at calibration neutron field. The average neutron energy at measured place was between 42 and 158 keV and higher than that of calibration field of 500 keV. So, the measured equivalent dose in nuclear power plant could be overestimated compared to the real equivalent dose.

• Journal of the Earthquake Engineering Society of Korea
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• v.5 no.5
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• pp.47-56
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• 2001

In order to assure the safety of NPP structures, margin of safety or conservatism is incorporated in each design step. Seismic risk evaluation of NPP structures is performed based on the realistic capacity and response of structure eliminated the safety margin and conservatism. In this study, the comparative study on the various evaluation methods of the inelastic energy absorption capacity was performed. The inelastic energy absorption capacity due to the nonlinear behavior of structures has significant effect on the results of seismic probabilistic risk assessment. And the comparison study of the HCLPF(high confidence of low probability of failure) values according to the inelastic energy absorption factors was performed. As a conclusion, the inelastic energy absorption factor of NPP containment structure is estimated about 1.5~1.75. It is essential to estimate the nonlinear behavior of structure and its ductility factor correctly for the seismic risk assessment.

• Journal of the Korean Society for Nondestructive Testing
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• v.36 no.3
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• pp.231-235
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• 2016

There are two methods for conducting the containment local leakage rate test (LLRT) in nuclear power plants: the make-up flow rate method and the pressure decay method. The make-up flow rate method is applied first in most power plants. In this method, the leakage rate is measured by checking the flow rate of the make-up flow. However, when it is difficult to maintain the test pressure because of excessive leakage, the pressure decay method can be used as a complementary method, as the leakage rates at pressures lower than normal can be measured using this method. We studied the method of measuring over leakage using the pressure decay method for conducting the LLRT for the containment building at a nuclear power plant. We performed experiments under conditions similar to those during an LLRT conducted on-site. We measured the characteristics of the leakage rate under varies pressure decay conditions, and calculated the compensation ratio based on these data.