• 한국전산구조공학회논문집
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• 제28권5호
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• pp.535-542
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• 2015

원자로 격납건물은 냉각재상실사고와 같이 내부의 과도한 압력이 유발되는 사고에 있어서도 방사성 물질이 외부로 누출되지 않도록 막는 최종의 방벽이다. 이러한 격납건물의 기능적 중요성에 기인하여, 건설 초기 구조건전성시험(SIT)을 수행한다. 이러한 SIT거동을 가장 실제와 가깝게 예측하기 위한 해석 연구를 수행하였다. 해당 연구의 결과는 2편의 논문으로 정리되었는데, 본 논문은 그 중 II편으로 I편의 해석모델 구성 시의 주요 고려사항의 분석 및 예비해석 결과를 반영한 상세 해석 모델의 구성 과정 및 해석 결과를 제시하고 있다. 특히 비부착식 텐던으로 시공된 구조물에서 덕트관에 의한 강성 저감효과 및 덕트관을 사이에 둔 텐던과 콘크리트간의 밀착 여부에 따른 영향을 해석 시 최대한 고려하고자 하였다. 이러한 과정을 통해 구축된 해석 모델에 따른 변위과 신고리 3호기 SIT 측정변위를 비교한 결과, ASME CC-6000 기준을 충분히 만족시키는 결과가 나타남을 확인하였다.

• 한국입자에어로졸학회지
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• 제10권4호
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• pp.137-154
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• 2014

문헌조사를 통해 그동안 선행연구로부터 밝혀진 방사성 에어로졸의 특성을 종합하고 정리하였다. 가상사고 실험 중 각재계통 및 원자로건물에서 측정한 에어로졸의 특성, 냉각재계통 및 원자로건물에서의 방사성 에어로졸 거동 해석을 위해 사용된 모델 에어로졸의 특성, 공학적 안전설비 성능평가를 위한 실험에 사용된 모델 에어로졸의 특성 등과 관련한 선행연구 내용을 종합해 볼 때, 원전사고 시 발생하는 에어로졸의 MMD는 $0.1{\sim}5{\mu}m$, GSD는 1.33~2.9, 에어로졸 농도는 $0.06{\sim}132g/m^3$의 범위를 보였다. 향후 공학적 안전설비의 설계를 위한 MMD와 GSD의 대표값은 대략 $1.5{\mu}m$와 1.8 내외라고 할 수 있으며, 에어로졸 농도는 대략 $10g/m^3$을 대표값으로 볼 수 있다.

• Nuclear Engineering and Technology
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• 제47권3호
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• pp.293-305
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• 2015

In this paper, we suggest the conceptual design of a water-cooled reactor system for a low-pressure inherent heat sink nuclear desalination plant (LIND) that applies the safety-related design concepts of high temperature gas-cooled reactors to a water-cooled reactor for inherent and passive safety features. Through a scoping analysis, we found that the current LIND design satisfied several essential thermal-hydraulic and neutronic design requirements. In a thermal-hydraulic analysis using an analytical method based on the Wooton-Epstein correlation, we checked the possibility of safely removing decay heat through the steel containment even if all the active safety systems failed. In a neutronic analysis using the Monte Carlo N-particle transport code, we estimated a cycle length of approximately 6 years under 200 $MW_{th}$ and 4.5% enrichment. The very long cycle length and simple safety features minimize the burdens from the operation, maintenance, and spent-fuel management, with a positive impact on the economic feasibility. Finally, because a nuclear reactor should not be directly coupled to a desalination system to prevent the leakage of radioactive material into the desalinated water, three types of intermediate systems were studied: a steam producing system, a hot water system, and an organic Rankine cycle system.

• 가스산업과 기술
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• 제5권1호
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• pp.39-47
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• 2002

With the expansion of natural gas demands in many countries, the necessity of LNG receiving terminals has been increased. The offshore LNG Floating Storage and Regasification Unit (FSRU) attracts attentions not only for a land based LNG receiving terminal alternative, but also for a feasible and economic solution. Nowadays, as the reliability of offshore oil and gas floating facilities and LNG carriers gains with proven worldwide operations, the FSRU can achieve a safety level that can be comparable to an onshore terminal. The design development related with safety features of the FSRU has been extensively carried out by oil and gas companies, shipyards, engineering companies, and equipment vendors, and has been successful so far in many fields. The construction of the FSRU can be achieved by integrating various technologies and experiences from many disciplines and many participating companies and vendors. In this paper, reviews on some of the important design features and design improvements on FSRU together with the practical construction aspects in cargo containment, vaporization system, ESD system, and operation modes, have been covered in comparison with actual LNG carrier, onshore receiving terminal, and FPSO systems. In order to materialize an FSRU project, the technical and economic justification has to be preceded. It is believed that once the safety and technical soundness is convinced, the FSRU can bring a higher project feasibility by reducing the overall construction time and cost. Through this study, an FSRU design readily applicable to an actual project has been developed by incorporating experiences gained from many marine and offshore projects. The wide use of proven standard technologies adopted in the series construction of LNG carriers and offshore FPSOs will bring the project efficiency and reliability.

• Nuclear Engineering and Technology
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• 제45권5호
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• pp.605-612
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• 2013

Removal of decay heat from an operating reactor during a prolonged station blackout condition is a big concern for reactor designers, especially after the recent Fukushima accident. In the case of a prolonged station blackout condition, heat removal is possible only by passive means since no pumps or active systems are available. Keeping this in mind, the AHWR has been designed with many passive safety features. One of them is a passive means of removing decay heat with the help of Isolation Condensers (ICs) which are submerged in a big water pool called the Gravity Driven Water Pool (GDWP). The ICs have many tubes in which the steam, generated by the reactor core due to the decay heat, flows and condenses by rejecting the heat into the water pool. After condensation, the condensate falls back into the steam drum of the reactor. The GDWP tank holds a large amount of water, about 8000 $m^3$, which is located at a higher elevation than the steam drum of the reactor in order to promote natural circulation. Due to the recent Fukushima type accidents, it has been a concern to understand and evaluate the capability of the ICs to remove decay heat for a prolonged period without escalating fuel sheath temperature. In view of this, an analysis has been performed for decay heat removal characteristics over several days of an AHWR by ICs. The computer code RELAP5/MOD3.2 was used for this purpose. Results indicate that the ICs can remove the decay heat for more than 10 days without causing any bulk boiling in the GDWP. After that, decay heat can be removed for more than 40 days by boiling off the pool inventory. The pressure inside the containment does not exceed the design pressure even after 10 days by condensation of steam generated from the GDWP on the walls of containment and on the Passive Containment Cooling System (PCCS) tubes. If venting is carried out after this period, the decay heat can be removed for more than 50 days without exceeding the design limits.