• Journal of Radiation Protection and Research
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• 제14권2호
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• pp.37-44
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• 1989

원자로 냉각재에서의 방사능 누적 저감을 위한 수질 관리 개선 방안으로 현재의 coordinated lithium-boron 운전 방식을 elevated lithium 방식으로 전환시켜 냉각재의 pH를 높게 유지시키는 기법에 대해 검토하였다. 국내 PWR원전에서의 pH와 원자로 냉각재내의 방사능 누적 관계를 분석하였으며 그 결과, 고 pH 운전이 현재의 pH 운전 방법보다는 방사능 누적 저감에 유리하다는 것을 알 수 있었다. 이러한 결과는 냉각재중의 부식생성물의 구성이 magnetite 보다는 nickel ferrite 쪽이 지배적인 비중을 차지하고 있음을 보여주는 것이며, 고 pH 운전 범위는 pH 7.0-7.4가 적합한 것으로 나타났다.

• 한국화재소방학회:학술대회논문집
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• 한국화재소방학회 2003년도 춘계학술논문발표회논문집
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• pp.259-264
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• 2003

The fire protection system of Nuclear Power Plants(NPPs) is an integrated system that is applied multi-field technology. So, it needs synthetic design and analysis, that is, the plan of fire protection, fire compartment, fire detection, fire suppression, and success of safety shut down, etc. In case of a fire in NPPs, secure the safety of reactor and minimize the radioactivity contamination. For this purpose, perform the fire risk analysis and make up the deducted problem through the improvement of design or the change of operation process.

• Nuclear Engineering and Technology
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• 제54권6호
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• pp.2135-2146
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• 2022

A nuclear energy facility is one of the most critical facilities to be safely protected during and after operation because the physical destruction of its barriers by an external attack could release radioactivity into the environment and can cause harmful effects. The barrier walls of nuclear energy facilities should be sufficiently robust to protect essential facilities from external attack or sabotage. Physical protection system (PPS) vulnerability assessment of a typical small nuclear research reactor was carried out by simulating an external attack with a tri-nitro toluene (TNT) shaped charge and results are presented. The reinforced concrete (RC) barrier wall of the research reactor located at a distance of 50 m from a TNT-shaped charge was the target of external attack. For the purpose of the impact assessment of the RC barrier wall, a finite element method (FEM) is utilized to simulate the destruction condition. The study results showed that a hole-size of diameter 342 mm at the front side and 364 mm at the back side was created on the RC barrier wall as a result of a 143.35 kg TNT-shaped charge. This aperture would be large enough to let at least one person can pass through at a time. For the purpose of the PPS vulnerability assessment, an Estimate of Adversary Sequence Interruption (EASI) model was used, which enabled the determination of most vulnerable path to the target with a probability of interruption equal to 0.43. The study showed that the RC barrier wall is vulnerable to a TNT-shaped charge impact, which could in turn reduce the effectiveness of the PPS.

• 생명과학회지
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• 제12권5호
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• pp.584-588
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• 2002

Apergillus niger LK의 epoxide hydrolase 활성을 이용하여 chiral styrene oxide를 제조할 수 있는 hollow-fiber 반응기 기반의 비대칭 분할 시스템을 개발하였다. 라세믹 styrene oxide 기질을 dodecane 유기용매에 용해시켜 hollow-fiber 반응기의 lumen 부위로 공급하였으며, 생촉매인 A. niger LK 미세분말은 shell 부위로 공급함으로써 막 표면에서 비대칭 분할 반응을 수행하였다. 반응 산물로 생성되는 phenyl-1,2-ethandiol에 의한 epoxide hydrolase 활성 저해효과를 감소시키기 위하여 2번째 hollow-fiber 반응기에서 완충용액을 이용하여 diol을 추출하여 제거시켰다. 2성분 용매를 사용한 cascade형 hollow-fiber 반응기 시스템을 이용하여 광학적으로 순수한 (ee > 99%) (5)-styrene oxide를 19.5% (이론 수율 대비 39%)의 수율로 얻을 수 있었다.

• Journal of Radiation Protection and Research
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• 제37권2호
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• pp.63-69
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• 2012

The second phase of the national program for fusion energy development in Korea starts from 2012 for design and construction of the fusion DEMO reactor. Radiological assessment for the fusion reactor is one of the key tasks to assure its licensability and the starting point of the assessment is determination of the source terms. As the first effort, the activities of the coolant due to activated corrosion product (ACP) were estimated. Data and experiences from fission reactors were used, in part, in the calculations of the ACP concentrations because of lack of operating experience for fusion reactors. The MCNPX code was used to determine neutron spectra and intensities at the coolant locations and the FISPACT code was used to estimate the ACP activities in the coolant of the fusion DEMO reactor. The calculated specific activities of the most nuclides in the fusion DEMO reactor coolant were 2-15 times lower than those in the PWR coolant, but the specific activities of $^{57}Co$ and $^{57}Ni$ were expected to be much higher than in the PWR coolant. The preliminary results of this study can be used to figure out the approximate radiological conditions and to establish a tentative set of radiological design criteria for the systems carrying coolant in the design phase of the fusion DEMO reactor.

• Nuclear Engineering and Technology
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• 제49권7호
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• pp.1442-1450
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• 2017

In liquid metal fast breeder reactors, postulated failures of the plant protection system may lead to serious unprotected accidental consequences. Unprotected transients are generically categorized as transient overpower accidents and transient under cooling accidents. In both cases, core meltdown may occur and this can lead to a molten fuel coolant interaction (MFCI). The understanding of MFCI phenomena is essential for study of debris coolability and characteristics during post-accident heat removal. Sodium is used as coolant in liquid metal fast breeder reactors. Viewing inside sodium at elevated temperature is impossible because of its opaqueness. In the present study, a methodology to depict MFCI phenomena using a flat panel detector based imaging system (i.e., real time radiography) is brought out using a woods metal-water experimental facility which simulates the $UO_2-Na$ interaction. The developed imaging system can capture attributes of the MFCI process like jet breakup length, jet front velocity, fragmented particle size, and a profile of the debris bed using digital image processing methods like image filtering, segmentation, and edge detection. This paper describes the MFCI process and developed imaging methodology to capture MFCI attributes which are directly related to the safe aspects of a sodium fast reactor.

• Nuclear Engineering and Technology
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• 제32권4호
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• pp.316-327
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• 2000

The Emergency Core Cooling System in current nuclear power plants typically has a considerable number of complex functions and largely cumbersome operator interfaces. Functions for initiation, switch-over between various phases of operation, interlocks, monitoring, and alarming are usually performed by relays and analog comparator logic which are difficult to maintain and test. To improve problems of an analog based ECC (Emergency Core Cooling) System, the trip computer for ECCS based on Dynamic Safety System (DSS) is implemented. The DSS is a computer based reactor protection system that has fail-safe nature and performs a dynamic self-testing. The most important feature of the DSS is the introduction of test signal that send the system into a tripped state. The test signals are interleaved with the plant signals to produce an output which switches between a tripped and health state. The dynamic operation is a key feature of the failsafe design of the system. In this work, a possible implementation of the DSS using PLC is presented for a CANDU Reactor. ECC System of the CANDU Reactor is selected as the reference system.

• 제어로봇시스템학회:학술대회논문집
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• 제어로봇시스템학회 2001년도 ICCAS
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• pp.40.2-40
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• 2001

The protection system of the system communication, nuclear reactor and chemical reactor are representative of reliable system. This reliable system must be designed based on reliability as well as concept of safety, which is a failed system go a way of safe. Reliable system is composed of part of data acquisition, calculator, communication with redundancy, and a voter is important factor of reliability. Because it is serially connected. This paper presents a Design and Analysis of a Voting Mechanism considering Safety for reliable system Using EPLD. In the case of digital implementation a coincidence logic (voter) of reliable system, it needs CPU and memory, so increase a number of units. Therefore the failure rate and cost are increased on contrary when it is designed EPLD or FPGA.

• Nuclear Engineering and Technology
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• 제41권6호
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• pp.813-820
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• 2009

A rapid pneumatic transfer system (PTS) for an instrumental neutron activation analysis (INAA) is developed as an automatic irradiation facility involving the measurement of a short half-life nuclide and a delayed neutron counting system. Three new PTS designs with improved functions were constructed at the HANARO research reactor in 2006. The new system is composed of a manual system and an automatic system for both an INAA and a delayed neutron activation analysis (DNAA). The design and basic conception of a modified PTS are described, and the functions of system operation and control, radiation protection and emissions of radioactive gas are improved. In addition, a form of capsule transportation of these systems is tested. The experimental results pertaining to the irradiation characteristics with variation of the neutron flux and the temperature of the irradiation position with the irradiation time are presented, as is an analysis of the reference material for analytical quality control and uncertainty assessments.

• Nuclear Engineering and Technology
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• 제27권2호
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• pp.165-175
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• 1995

가압경수로의 과압보호계통은 가장 심각한 비정상 과도운전시 원자로냉각재계통의 압력을 설계압력의 110% 이내로 유지시킬 수 있는 충분한 용량으로 설계되어져야 한다. 본 연구에서는 ABB-CE 설계의 2825 MWt 가압경수로에 기존의 스프링 탑재형 가압기 안전밸브 대신 SEBIM-POSRV를 채택할 경우 과압보호 기능 수행의 가능성을 연구하였다. 과압보호 기능을 수행하기 위한 SEBIM POSRV의 크기 및 작동 설정치를 영광 3, 4호기의 과압보호 해석에 사용했던 LTC 전산코드를 이용한 분석을 통해서 결정했다. 분석 결과 monobloc SEBIM POSRV를 이용한 과압보호계통은 원자로냉각재계통의 압력을 설계 압력의 110% 이내로 유지시킴으로써 ABB-CE 형태의 2825 MWt급 가압경수로에서 과압보호 기능을 수행할 수 있음이 입증되었다.