• Journal of Radiation Protection and Research
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• 제43권4호
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• pp.154-159
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• 2018

Background: To prevent small leakage accidents, a real-time and direct detection system for small leaks with a detection limit below that of existing systems, e.g. $0.5gpm{\cdot}hr^{-1}$, is required. In this study, a small-size beta detector, which can be installed inside the reactor containment (CT) building and detect small leaks directly, was suggested and its feasibility was evaluated using MCNPX simulation. Materials and Methods: A target nuclide was selected through analysis of radiation from radionuclides in the reactor coolant system (RCS) and the spectrum was obtained via a silicon detector simulated in MCNPX. A window was designed to reduce the background signal caused by other nuclides. The sensitivity of the detector was also estimated, and its shielding designed for installation inside the reactor CT. Results and Discussion: The beta and gamma spectrum of the silicon detector showed a negligible gamma signal but it also contained an undesired peak at 0.22 MeV due to other nuclides, not the $^{16}N$ target nuclide. Window to remove the peak was derived as 0.4 mm for beryllium. The sensitivity of silicon beta detector with a beryllium window of 1.7 mm thickness was derived as $5.172{\times}10^{-6}{\mu}Ci{\cdot}cc^{-1}$. In addition, the specification of the shielding was evaluated through simulations, and the results showed that the integrity of the silicon detector can be maintained with lead shielding of 3 cm (<15 kg). This is a very small amount compared to the specifications of the lead shielding (600 kg) required for installation of $^{16}N$ gamma detector in inside reactor CT, it was determined that beta detector would have a distinct advantage in terms of miniaturization. Conclusion: The feasibility of the beta detector was evaluated for installation inside the reactor CT to detect small leaks below $0.5gpm{\cdot}hr^{-1}$. In future, the design will be optimized on specific data.

• Nuclear Engineering and Technology
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• 제54권6호
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• pp.2077-2083
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• 2022

In the unlikely event of core disruptive accident in sodium cooled fast reactors, the reactor containment building would be bottled up with sodium and fission product aerosols. The behavior of these aerosols is crucial to estimate the in-containment source term as a part of nuclear reactor safety analysis. In this work, the evolution of sodium aerosol characteristics (mass concentration and size) is simulated using HAARM-S code. The code is based on the method of moments to solve the integro-differential equation. The code is updated to FORTRAN-77 and run in Microsoft FORTRAN PowerStation 4.0 (on Desktop). The sodium aerosol characteristics simulated by HAARM-S code are compared with the measured values at Aerosol Test Facility. The maximum deviation between measured and simulated mass concentrations is 30% at initial period (up to 60 min) and around 50% in the later period. In addition, the influence of humidity on aerosol size growth for two different aerosol mass concentrations is studied. The measured and simulated growth factors of aerosol size (ratio of saturated size to initial size) are found to be matched at reasonable extent. Since sodium is highly reactive with atmospheric constituents, the aerosol growth factor depends on the hygroscopic growth, chemical transformation and density variations besides coagulation. Further, there is a scope for the improvement of the code to estimate the aerosol dynamics in confined environment.

• Nuclear Engineering and Technology
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• 제56권1호
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• pp.132-140
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• 2024

In the event of a severe accident in Sodium Cooled Fast Reactors (SFR), the sodium combustion aerosols along with fission product aerosols would migrate to the environment through leak paths of the Reactor Containment Building (RCB) concrete wall under positive pressure. Understanding the characteristics of sodium aerosol transport through concrete leak paths is important as it governs the environmental source term. In this context, experiments are conducted to study the influence of various parameters like pressure, initial mass concentration, leak path diameter, humidity etc., on the transport and deposition of sodium aerosols in straight leak paths of concrete. The leak paths in concrete specimens are prepared by casting and the diameter of the leak path is measured using thermography technique. Aerosol transport experiments are conducted to measure the transported and plugged aerosol mass in the leak paths and corresponding plugging times. The values of differential pressure, aerosol concentration and relative humidity taken for the study are in the ranges 10-15 kPa, 0.65-3.04 g/m³ and 30-90% respectively. These observations are numerically simulated using 1-Dimensional transport equation. The simulated values are compared with the experimental results and reasonable agreement among them is observed. From the safety assessment view of reactor, the approach presented here is conservative as it is with straight leak paths.

• 한국전산구조공학회:학술대회논문집
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• 한국전산구조공학회 2009년도 정기 학술대회
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• pp.287-290
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• 2009

이 논문에서는 가압중수형(Pressurized Heavy Water Reactor) 프리스트레스 콘크리트 격납건물의 1/4 축소모델에 대한 극한내압능력과 전반적인 비선형거동에 관한 유한요소 해석을 수행하였다. 가압중수형 격납건물은 원통형 벽체와 돔으로 구성되었고, 4개의 부벽을 갖는다. 유한요소해석을 위해서 상용코드 ABAQUS를 이용하였고, 콘크리트, 철근 및 텐던에 대한 수치모델링을 작성하여 자중과 내압하중을 적용하였고, 텐던의 2% 변형률을 기준으로 극한내압능력을 평가하였다. 이때 사용된 재료모델로 콘크리트는 Concrete Damaged Plasticity 모델을 사용하였고, 철근과 텐던은 Elasto-Plastic 모델을 적용하였다. 유한요소 해석결과 콘크리트의 초기균열 0.41MPa에서 발생하였고, 극한내압은 0.56MPa 정도로 평가되었다.

• Nuclear Engineering and Technology
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• 제38권5호
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• pp.443-448
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• 2006

The OPAL (Open Pool Australian Light-water) reactor is currently being constructed to replace HIFAR (HI-Flux Australian Reactor, commissioned in 1958) in mid-2006. HIFAR will be shutdown for decommissioning after several months of simultaneous operation with OPAL for smooth transition of operating systems and business. OPAL is a 20 MW multipurpose research reactor for radioisotope production, irradiation services and neutron beam research. The OPAL reactor uses low enriched uranium fuel in a compact core, cooled by light water and moderated by heavy water, yielding maximum thermal flux not less than $4{\times}10^{14}ncm^{-2}s^{-1}$. The reactor containment building is constructed of reinforced concrete and has been designed to protect the reactor from all external events such as seismic occurrences and impact from a hypothetical light aircraft crash. This paper describes the main elements of the reactor design and its applications.

• 한국안전학회지
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• 제23권2호
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• pp.21-29
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• 2008

To evaluate the structural integrity of the NPP containment building more rigorously, the effective prestress, which is one of the most affecting elements, needs to be estimated exactly. This paper presents the results of an experimental study to determine the effective prestress force in prestressed concrete beams. It is possible to improve the effective prestress measuring method by test beam, which is being applied for the investigation of the nuclear power plant in operation. If experimentally evaluated Lift-Off method in this study can be coupled with test beam test currently being used in in-service nuclear power plant, it is possible to measure prestress loss of the tendon and the level of the effective prestress load.

• 제어로봇시스템학회:학술대회논문집
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• 제어로봇시스템학회 1996년도 한국자동제어학술회의논문집(국내학술편); 포항공과대학교, 포항; 24-26 Oct. 1996
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• pp.1519-1522
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• 1996

In this paper, we have developed a tele-robotic system for nozzle dam installation/removal works and tube relating maintenance works inside unclear power plant steam generator. Developed tele-robotic system consists of many hardwares including robot and a control system. Based on the 3 dimensional graphic simulation, a 6 D.O.F. hydraulic actuated robot and a 2 D.O.F. robot install/removal device have been developed. And also we deviced special tools for nozzle dam carry and bolting. For the tele-robot and other devices to be controlled at the nonradioactive area outside reactor containment building, we developed a tele-robot control system consisting of supervisory controller and remote controller.

• 전기학회논문지
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• 제59권5호
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• pp.996-1000
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• 2010

In order to apply VLC (visible light communication) in harsh environment of nuclear power plant in-containment building, the high luminance LEDs, which are key components of the VLC system, have been gamma irradiated at the dose rate of 4 kGy/h during 72 hours up to a total dose of 288 kGy. The radiation induced coloration effect in the high luminance LED bulb made of acryl or plastic material was observed. In the VLC wireless data transmission experiment using the high luminance LEDs irradiated by high dose rate gamma-ray, the radiation induced coloration effect of the high luminance LED bulb extended the communication distance compared to non-irradiated LEDs.

• 한국지진공학회:학술대회논문집
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• 한국지진공학회 1997년도 춘계 학술발표회 논문집 Proceedings of EESK Conference-Fall 1997
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• pp.225-232
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• 1997

A study of effects of design parameters on the seismic responses of base isolated structure is performed to reduce the seismic responses using a linear tw0-degree of freedom system and a lumped-mass model of a nuclear power p;ant(NPP). From the simplified 2 DOF system the optimal isolation frequency being less than 1/10th of the fundamental frequency of superstructure is obtained, and the isolator damping minimizing the peak acceleration depends on superstructure frequency. From the time history analyses for lumped mass model of NPP the optimal damping is calculated as 40% in containment building and 65% in reactor internal structure. Similar results are obtained in 2 DOF system

• Nuclear Engineering and Technology
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• 제42권3호
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• pp.313-318
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• 2010

Alnine pellets were installed in a nuclear power plant for one or two operation cycles and measured by electron spin resonance (ESR) spectrometers for dosimetry. Dose and "x/y ratio", i.e., satellite peak over main center peak ratio, were measured for the returned alanine dosimeters from the nuclear power plant and compared to the values of reference alanine dosimeters exposed only to gamma rays. The variation of the x/y ratio change depended on the population of radicals from each radiation component with different LET. The gamma dose in a mixed radiation field was estimated by an additive gamma ray irradiation experiment and the measured dose rate at specified locations in the containment building.