• 한국가스학회:학술대회논문집
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• 한국가스학회 2001년도 추계학술발표회 논문집
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• pp.122-129
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• 2001

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2005년도 추계학술발표회
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• pp.801-802
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• 2005

• 한국안전학회지
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• 제33권6호
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• pp.136-143
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• 2018

In this study, we conducted a pilot study on the multiple spurious operations (MSO) analysis in the fire probabilistic safety assessment (PSA) of domestic nuclear power plant (NPP) to identify the degree of influence of the operator actions used in the MSO mitigation strategies. The MSO scenario of the domestic reference NPP selected for this study is refueling water tank (RWT) drain down event. It could be caused by spurious operations of the containment spray system (CSS) of the reference NPP. The RWT drain down event can be stopped by the main control room (MCR) operator actions for stopping the operation of CSS pump or closing the CSS motor operated valve if the containment spray actuation signal (CSAS) is spuriously actuated. Outside the MCR, it can be stopped by operator actions for closing the CSS manual valves or motor operated valve or stopping the operation of CSS pump. The quantification result of a fire PSA model that takes into account all recovery actions for the RWT drain down event lead to risk reduction by about 95%, compared with quantification result of fire PSA model without considering them. Among the various operator actions, the recovery action for the spurious CSAS operations and the operator action for the manual valve are identified as the most important operator actions. This study quantitatively showed the extent to which the operator actions used as MSO countermeasures have affected the fire PSA quantification results. In addition, we can see the rank of importance among the operator recovery actions in quantitative terms.

• 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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• 제54권9호
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• pp.3310-3316
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• 2022

Owing to the rising concerns about the safety of nuclear power plants (NPP), it is essential to study the venturi scrubber in detail, which is a key component of the filtered containment venting system (FCVS). FCVS alleviates the pressurein containment byfiltering and venting out the contaminated air. Themain objective of this research was to perform a CFD investigation of different configurations of a circular, non-submerged, self-priming venturi scrubber to estimate and improve the performance in the removal of elemental iodine from the air. For benchmarking, a mass transfer model which is based on two-film theory was selected and validated by experimental data where an alkaline solution was considered as the scrubbing solution. This mass transfer model was modified and implemented on a unique formation of two self-priming venturi scrubbers in series. Euler-Euler method was used for two-phase modeling and the realizable K-ε model was used for capturing the turbulence. The obtained results showed a remarkable improvement in the removal of radioactive iodine from the air using a series combination of venturi scrubbers. The removal efficiency was improved at every single data point.

• Nuclear Engineering and Technology
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• 제41권8호
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• pp.1005-1014
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• 2009

The Dual Unit Optimizer 2000 MWe (DUO2000) is put forward as a new design concept for large power nuclear plants to cope with economic and safety challenges facing the $21^{st}$ century green and sustainable energy industry. DUO2000 is home to two nuclear steam supply systems (NSSSs) of the Optimized Power Reactor 1000 MWe (OPR1000)-like pressurized water reactor (PWR) in single containment so as to double the capacity of the plant. The idea behind DUO may as well be extended to combining any number of NSSSs of PWRs or pressurized heavy water reactors (PHWRs), or even boiling water reactors (BWRs). Once proven in water reactors, the technology may even be expanded to gas cooled, liquid metal cooled, and molten salt cooled reactors. With its in-vessel retention external reactor vessel cooling (IVR-ERVC) as severe accident management strategy, DUO can not only put the single most querulous PWR safety issue to an end, but also pave the way to very promising large power capacity while dispensing with the huge redesigning cost for Generation III+ nuclear systems. Five prototypes are presented for the DUO2000, and their respective advantages and drawbacks are considered. The strengths include, but are not necessarily limited to, reducing the cost of construction by decreasing the number of containment buildings from two to one, minimizing the cost of NSSS and control systems by sharing between the dual units, and lessening the maintenance cost by uniting the NSSS, just to name the few. The latent threats are discussed as well.

• Nuclear Engineering and Technology
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• 제48권2호
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• pp.505-517
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• 2016

The structural integrity of a dual-purpose metal cask currently under development by the Korea Radioactive Waste Agency (KORAD) was evaluated, through numerical simulations and a model test, under high-speed missile impact reflecting targeted aircraft crash conditions. The impact conditions were carefully chosen through a survey on accident cases and recommendations from literature. In the impact scenario, a missile flying horizontally hits the top side of the cask, which is freestanding on a concrete pad, with a velocity of 150 m/s. A simplified missile simulating a commercial aircraft engine was designed from an impact loade-time function available in literature. In the analyses, the dynamic behavior of the metal cask and the integrity of the containment boundary were assessed. The simulation results were compared with the test results for a 1:3 scale model. Although the dynamic behavior of the cask in the model test did not match exactly with the prediction from the numerical simulation, other structural responses, such as the acceleration and strain history during the impact, showed very good agreement. Moreover, the containment function of the cask survived the missile impact as expected from the numerical simulation. Thus, the procedure and methodology adopted in the structural numerical analyses were successfully validated.

• Nuclear Engineering and Technology
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• 제42권1호
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• pp.1-8
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• 2010

Geological disposal is designed to provide safe containment of radioactive waste for very long times, with the containment provided by a combination of engineered and geological barriers. In the extreme long term, after many thousands of years or longer, residual amounts of long-lived radionulides such as Cl-36, but also radionuclides in the natural decay chains, may be released into the environment normally accessed and used by humans, termed here, the biosphere. It is necessary to ensure that any such releases meet radiation protection objectives through the development of a safety case, which will include assessment of radiation doses to humans. The design of such dose calculations over such long timeframes is not straightforward, because of the range of potentially relevant assumptions which could be made, concerning environmental change and changes in human behavior. These conceptual uncertainties are additional to those that more typically arise, for example, in the assessment of present day situations, but which also have to be addressed. The issue has therefore been subject to international cooperation for many years. This paper summarizes the evolution and results of that collaboration leading up to the present day, taking account of developments in international recommendations on radiation protection objectives and the more recent greater focus on preparation of site specific safety cases.

• 한국가스학회지
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• 제9권4호
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• pp.57-61
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• 2005

이 연구는 결함수분석(FTA)를 이용해 완전방호 형식 및 멤브레인 형식의 KOGAS 탱크 설계에 대한 비교 정량적 위험성평가를 위해 수행하였다. 이 연구를 위해 표준 완전방호식 탱크와 초기 멤브레인 탱크 및 4가지의 개선된 멤브레인 탱크를 포함한 총 6개 모델에 대해 위험성을 평가하였다. 이 연구에서는 FTA를 이용해 누출빈도를 정량화 하였다. 분석의 명확성 및 일관성을 위해서 모든 경우는 동일한 고장수(fault tree)를 이용해 정량화하였다. 개선되지 않은 멤브레인 저장탱크(초기 모델)를 제외하고 예측된 위험도 수준은 매우 유사해서 각각의 탱크는 동일한 위험도 수준(the same level of risk)을 나타내는 것으로 평가되었다. 펌프 낙하에 의한 손상은 완전방호식 탱크에 비해서 박판으로 되어있는 멤브레인 탱크가 두드러지게 큰 것으로 예측되었다.

• 한국가스학회지
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• 제12권2호
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• pp.18-24
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• 2008

본 논문은 완전방호식 LNG 저장탱크의 9% 니켈강재 내부탱크에 대해 응력 및 변형거동 특성을 FEM으로 해석하였다. 내부탱크의 강도안전성을 확보하기 위해 기존 LNG 저장탱크에 유체정압, 초저온하중 등을 가할 경우, 취약부로 알려진 지역에 인장강선을 설치하였다. 기존의 9% 니켈강재 내부탱크와 내부탱크의 하단부에 인장강선을 설치한 탱크에 대한 FEM 해석결과에 의하면, 새롭게 설계한 내부탱크는 인장강선을 설치하지 않은 기존의 내부탱크에 비해 더 안전한 것으로 나타났다. FEM 해석결과에 따르면, 내부탱크의 강도안전성 향상을 위해 내부탱크의 하단부에 직경 50 mm를 갖는 $3{\sim}4$개의 인장강선을 설치할 것을 권장하였는데, 이것은 결국 내부탱크의 응력 및 변위량을 낮출 수 있다. 따라서, 내부탱크에 인장강선을 설치하는 것은 LNG 저장탱크의 강도안전성 확보를 위해 기존 내부탱크에서 도입한 스티프너와 톱거더를 대체한 또 다른 안전장치로 활용될 수 있다.