• Proceedings of the Korean Institute of Navigation and Port Research Conference
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• 2011.06a
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• pp.343-345
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• 2011

대형 여객선은, "a ship is its own best lifeboat"라는 개념을 바탕으로 여객선의 안전성(survivability) 향상을 위한 설계가 요구되고 있으며, 이를 위하여, 사고가 발생하더라도 선박의 자체 추진력으로 안전하게 항구까지 귀항하여야 한다는, 여객선의 안전귀항(SRtP) 이라는 개념을 IMO SOLAS에 적용시켰다. SOLAS의 여객선 안전귀항 관련 조항은, 길이 120m 이상인 선박 또는 3개 이상의 주 수직격벽을 가진 선박으로서 2010년 07월 01일 이후 건조되는 여객선에 적용된다. 여객선 안전귀항 관련 조항은 화재와 침수사고에 적용되며, 사고분계점을 넘지 아니하는 사고가 발생할 경우 자체 추진력으로 여객선의 안전한 귀항을 위하여 사용 가능한 상태로 유지되어야 하는 시스템들에 대한 설계 기준, 사고분계점을 초과하는 화재 사고가 발생하였을 경우 질서 정연한 탈출 및 퇴선을 지원하기 위하여 작동상태의 유지가 요구되는 시스템 설계 기준, 사고분계점에 대한 정의, 사고 발생 후에도 여객 및 승무원의 건강을 유지 확보하기 위한 안전구역에 대한 기준들을 요구하고 있다. 본 연구에서는, 여객선 안전귀항 관련 법규들을 검토하고, 여객선 안전귀항을 위한 시스템들의 능력 평가 방법과 안전귀항 관련 조항 만족을 위한 시스템들의 요구사항들을 검토하였다.

• Proceedings of the Korea Institute of Fire Science and Engineering Conference
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• 2013.11a
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• pp.193-194
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• 2013

후쿠시마 원전 사고 후 극한자연재해 및 광역재해에 대한 관심이 높아지고 있다. 후쿠시마 원전사고처럼 설계기준을 초과하는 극한재해가 발생하면 원전의 안전설비가 이용불능이 될 수 있고, 이로 인하여 중대사고가 발생할 수 있다. 이와 같은 중대사고로 진행되는 것을 방지하기 위하여 국내원전에서는 대체수단으로서 이동형 비상설비를 각 원전에 배치하고 있다. 본 논문에서는 이동형 비상설비를 활용하기 위한 비상대체설비운영지침서 개발 내용을 소개하였다.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.15 no.2
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• pp.64-70
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• 2019

In recent years, attention has been paid to the integrity of steam generator (SG) tubes due to severe accident and beyond design basis accident conditions. In these transient conditions, steam generator tubes may be damaged by high temperature and pressure, which might result in a risk of fission products being released to the environment due to the failure. Alloy 690 which has increased the Cr content has been replaced for the SG tube due to its high corrosion resistance against stress corrosion cracking (SCC). However, there is lack of research on the high temperature creep rupture and life prediction model of Alloy 690. In this study, creep test was performed to estimate the high temperature creep rupture life of Alloy 690 using tube specimens. Based on manufacturer's creep data and creep test results performed in this study, creep life prediction was carried out using the Larson-Miller (LM) Parameter, Orr-Sherby-Dorn (OSD) parameter, Manson-Haford (MH) parameter, and Wilshire's approach. And a hyperbolic sine (sinh) function to determine master curves in LM, OSD and MH parameter methods was used for improving the creep life estimation of Alloy 690 material.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.15 no.4
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• pp.391-402
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• 2017

The Korea Radioactive Waste Agency (KORAD) has developed a concrete cask for the dry storage of spent nuclear fuel that has been generated by domestic light-water reactors. During long-term storage of spent nuclear fuel in concrete casks kept in dry conditions, the integrity of the concrete cask and spent nuclear fuel must be maintained. In addition, the radiation dose rate must not exceed the storage facility's design standards. A suitable shielding design for radiation protection must be in place for the dry storage facilities of spent nuclear fuel under normal and accident conditions. Evaluation results show that the appropriate distance to the annual dose rate of 0.25 mSv for ordinary citizens is approximately 230 m. For a $2{\times}10$ arrangement within storage facilities, rollover accidents are assumed to have occurred while transferring one additional storage cask, with the bottom of the cask facing the controlled area boundary. The dose rates of 12.81 and 1.28 mSv were calculated at 100 m and 230 m from the outermost cask in the $2{\times}10$ arrangement. Therefore, a spent nuclear fuel concrete cask and storage facilities maintain radiological safety if the distance to the appropriately assessed controlled area boundary is ensured. In the future, the results of this study will be useful for the design and operation of nuclear power plant on-site storage or intermediate storage facilities based on the spent fuel management strategy.

• Journal of the Korean Society of Marine Environment & Safety
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• v.17 no.3
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• pp.265-268
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• 2011

In this paper, we designed and fabricated that fog alarm equipment detected to crew in maritime fog. Developed fog alarm equipment was combined sensor and a transmitter-receiver module using emitting device unit and receiver photoelectric element infrared LED using only the receive sensitivity was low, miniaturization. Experiment of the fabricated device had a standard that was humidity 70%, the fabricated one generating artificial-fog within visibility 1km. When humidity is over 70%, the fabricated one generates alarming sounds for a warning. When developed device apply to vessel will be able to respond quickly, according to dense fog in the accident.

• Nuclear Engineering and Technology
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• v.27 no.4
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• pp.470-482
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• 1995

The Ulchin 3&4, which are 2825 MWt PWRs, adopted Safety Depressurization System (SDS) to mitigate the beyond design basis event of Total Less of Feedwater(TLOFW). In this study the results and methodology of the analyses for the determination of SDS bleed capacity are discussed. The SDS design bleed capacity has been determined from the CEFLASH-4AS/REM simulation according to the following design criteria : 1) Each SDS flow path, in conjunction with one of two High Pressure Safety Injection (HPSI) pumps, is designed to have a sufficient capacity to prevent core uncovery if one SDS path is opened simultaneously with the opening of the Pressurizer Safety Valves (PSVs). 2) Both SDS bleed paths are designed to have sufficient total capacity with both HPSI pumps operating to prevent core uncovery if the Feed and Bleed (F&B) initiation is delayed up to thirty minutes from the time of the PSVs lift. To verify the results of CEFLASH-4AS/REM simulation a comparative analysis kas also been per-formed by more sophisticated computer code, RELAP5/MOD3. The TLOFW event without operator recovery and TLOFW event with F&B are analyzed. The predictions by the CEFLASH-4AS/REM of the transient too phase system behavior are in good qualitative and quantitative agreement with those by the RELAP5/MOD3 simulation. Both of the results of analyses by CEFLASH-4AS/REM and RELAP5/MOD3 have demonstrated that decay heat removal and core inventory make-up can be successfully accomplished by F&B operation during now event for the Ulchin 3&4.

• Journal of Korean Tunnelling and Underground Space Association
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• v.22 no.3
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• pp.293-310
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• 2020

The purpose of this study is to present the most effective smoke exhaust mode by comparing the quantitatively evaluated risks according to the smoke exhaust mode when a train fire occurs in a subway platform. Therefore, applying the typical subway platform as a model, train fire scenarios are developed with the evacuation start time and location of the fire train for each exhaust mode. The fire accident rates (F) are calculated and the number of fatalities (N) was quantitatively estimated by fire analysis and evacuation analysis for each scenario. In addition, the F/N curve compared with the social risk assessment criteria and the following conclusions were obtained. In the event of a train fire at the subway station platform, the evacuation must start up within 600 s in maximum to ensure the evacuees' safety. To secure evacuation safety, it is advantageous to operate the HVAC system of the platform in the air-supply mode at station without TVF. Comparing the F/N curve for each exhaust mode with the social risk criteria, it turned out that the risk significantly exceeds the social risk criteria in case of no mechanical ventilation. As a result, this paper shows that the ventilation mode in which TVF are exhausted and HVAC system is operated in the pressurized mode are the most effective smoke exhaust mode for ensuring evacuation safety.

• Journal of Radiation Protection and Research
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• v.40 no.3
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• pp.162-167
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• 2015

Although the perfomance indicators of the nuclear power plants in Korea show optimal, it requires detailed analysis and discussion centered on the radiation dose. As analysis methods, analysis on the radiation dose of nuclear power plants over the past five years was assessed by comparing the relevant radiation dose of radiation workers and per capita average annual radiation dose of the world's major nuclear power stations was also analyzed. The radiation workers over the annual radiation dose limit of 50 mSv were not. The contrast ratio of the radiation exposure according to the reactor type was the normal operation of PHWR was 6.2% higher than those of the PWR. This shows the radiation work of PHWR during normal driving operation is much more than those of PWR. According to the Performance Indicators of the World Association of Nuclear Operator, the annual radiation dose per unit in 2013 showed 527 man-mSv of Korea is the best country among the major nuclear power generating states, the world average was 725 man-mSv. The annual per capita radiation dose is about 80% less than 1 mSv of the public dose limit and also the average per capita dose showed a very low level as 0.82 mSv. Workers in related organizations showed 1.07 mSv, the non-destructive inspection agency workers showed 3.87 mSv. The remarkable results were due to radiation reduced program such as development of radiation shielding and radiation protection. In conclusion, the radiation exposured dose of nuclear power plants workers in Korea showed a trend which is ideally reduced. But more are expected to be difficul and the psychological insecurity against the operation of the nuclear power plants is existed to the residents near the nuclear power plants. So the radiation dose reduction policy and radiation dose follow up study of nuclear power plants will be continously excuted.