• 대한기계학회논문집A
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• 제38권2호
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• pp.157-166
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• 2014

경수로 핵연료가 원자로내에서 연소되는 동안 핵연료 펠릿에서부터 피복관까지 온도해석은 핵연료 안전 해석에 있어 중요한 요소이며, 경수로 핵연료 온도 해석을 하기 위해서는 간극 모델 개발이 필수적이다. 간극 열전도도는 특성상 간극 두께값에 의존적이게 되며 이러한 특성으로 인해 다차원 간극 열전도도 모델이 비선형적 거동을 보인다. 본 연구에서는 선형화된 다차원 간극 열전도도 모델 개발을 위해 가상 연결 간극 요소를 제안하였다. 제안된 간극 연결 요소에 간극 열전도도를 적용하기 위해 등가 열전달 계수를 정의하였다. 제안된 모듈을 평가하기 위해 상용코드 ANSYS APDL 을 이용하여 열-구조 연계 해석 모듈을 구현하였으며, 다양한 예제를 통해 정확성과 수렴성을 평가하였다.

• 한국산학기술학회:학술대회논문집
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• 한국산학기술학회 2003년도 춘계학술발표논문집
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• pp.350-352
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• 2003

In this study plastic optical fibers(POFs) were considered as light-transmitting media and substrates for the potential use in photocatalytic environmental purification system. After the characteristics of POFs in terms of light transmittance and absorption were determined at the beginning, the detailed investigation was further performed through the photocatalytic degradation of trichloroethylene(TCE), iso-propanol and etc. with TiO$_2$-coated optical fiber reactor systems(POFR). It is concluded that the use of POfs is preferred to quartz optical fibers(QOFs) since the advantages such as ease of handling, lower cost, relatively reasonable light attenuation at the wavelength of near 400nm can be obtained. Various geometrical reactor shapes have been constructed and applied for the last one and half years. For the use of POF in water phase treatment, however, more detailed scientific and engineering aspects should be envisaged. This case requires a suitable mixture to obtain more stable and innocuous immobilization of photocatalyst on POF.

• 대한기계학회논문집B
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• 제36권7호
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• pp.689-695
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• 2012

지지격자가 있는 봉다발과 축방향으로 평행한 유동에서, 봉다발 마찰계수와 지지격자 손실계수를 평가하였다. 시험부는 외경 9.5 mm, 길이 2000 mm 인 봉 25 개를 $5{\times}5$ 정사각 구조로 배열하여 제작하였으며, 봉 중심간 거리와 봉 외경의 비는 1.35 였다. 지지격자로는 plain 지지격자, split-vane 지지격자, hybrid-vane 지지격자를 이용하였다. 지지격자가 없는 봉다발의 마찰계수는 기존 상관식과 비교적 잘 일치하였다. 지지격자가 있는 봉다발 실험의 경우, hybrid-vane 지지격자에서 봉다발 마찰계수 및 지지격자 손실계수가 가장 크게 측정되었으며, 이는 지지격자의 유동단면 막음비 증가와 혼합날개 형상에 의한 유동 교란이 증가되기 때문인 것으로 판단된다. Re=$5{\times}10^5$ 조건에서 plain 지지격자, split-vane 지지격자, hybrid-vane 지지격자의 손실계수는 약 0.79, 0.80, 0.88 로 예측되었다.

• Nuclear Engineering and Technology
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• 제52권9호
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• pp.2130-2137
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• 2020

The standard method for calculating anti-neutrino emissions from a reactor involves knowing the fractional fission rates for the most important fissioning nuclides in the reactor. To calculate these rates requires detailed reactor physics calculations based upon the reactor design, fuel design, burnup dependent fuel composition, location of specific fuel assemblies in the core and detailed operational data from the reactor. This has only been published for a few reactors during specific time periods, whereas to be of practical use for anti-neutrino reactor monitoring it is necessary to be able to predict these on the publicly available information from any reactor, especially if using these data to subtract the anti-neutrino signal from other reactors to identify an undeclared reactor and monitor its operation. This paper proposes a method to estimate the fission fractions for a specific reactor based upon publicly available information and provides a database based upon a series of spent fuel inventory calculations using the FISPIN10 code and its associated data libraries.

• Nuclear Engineering and Technology
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• 제41권7호
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• pp.929-944
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• 2009

The present study is concerned with the extension of the Effective Convectivity Model (ECM) to the phase-change problem to simulate the dynamics of the melt pool formation in a Light Water Reactor (LWR) lower plenum during hypothetical severe accident progression. The ECM uses heat transfer characteristic velocities to describe turbulent natural convection of a melt pool. The simple approach of the ECM method allows implementing different models of the characteristic velocity in a mushy zone for non-eutectic mixtures. The Phase-change ECM (PECM) was examined using three models of the characteristic velocities in a mushy zone and its performance was compared. The PECM was validated using a dual-tier approach, namely validations against existing experimental data (the SIMECO experiment) and validations against results obtained from Computational Fluid Dynamics (CFD) simulations. The results predicted by the PECM implementing the linear dependency of mushy-zone characteristic velocity on fluid fraction are well agreed with the experimental correlation and CFD simulation results. The PECM was applied to simulation of melt pool formation heat transfer in a Pressurized Water Reactor (PWR) and Boiling Water Reactor (BWR) lower plenum. The study suggests that the PECM is an adequate and effective tool to compute the dynamics of core melt pool formation.

• Nuclear Engineering and Technology
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• 제48권1호
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• pp.268-273
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• 2016

An innovative safety concept for a light water reactor has been developed at the Korea Atomic Energy Research Institute. It is a unique concept that adopts both a fast heat transfer mechanism for a small containment and a changing mechanism of the cooling geometry to take advantage of the potential, thermal, and dynamic energies of the cold water in the containment. It can bring about rapid cooling of the containment and long-term cooling of the decay heat. By virtue of this innovative concept, nuclear fuel damage events can be prevented. The ultimate heat transfer mechanism contributes to minimization of the heat exchanger size and containment volume. A small containment can ensure the underground construction, which can use river or seawater as an ultimate heat sink. The changing mechanism of the cooling geometry simplifies several safety systems and unifies diverse functions. Simplicity of the present safety system does not require any operator actions during events or accidents. Therefore, the unique safety concept of PX can realize both economic competitiveness and inherent safety.

• Nuclear Engineering and Technology
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• 제49권6호
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• pp.1219-1225
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• 2017

In this work, a scalable algorithm for model calibration in nuclear engineering applications is presented and tested. The algorithm relies on the construction of surrogate models to replace the original model within the region of interest. These surrogate models can be constructed efficiently via reduced order modeling and subspace analysis. Once constructed, these surrogate models can be used to perform computationally expensive mathematical analyses. This work proposes a surrogate based model calibration algorithm. The proposed algorithm is used to calibrate various neutronics and thermal-hydraulics parameters. The virtual environment for reactor applications-core simulator (VERA-CS) is used to simulate a three-dimensional core depletion problem. The proposed algorithm is then used to construct a reduced order model (a surrogate) which is then used in a Bayesian approach to calibrate the neutronics and thermal-hydraulics parameters. The algorithm is tested and the benefits of data assimilation and calibration are highlighted in an uncertainty quantification study and requantification after the calibration process. Results showed that the proposed algorithm could help to reduce the uncertainty in key reactor attributes based on experimental and operational data.

• 유체기계공업학회:학술대회논문집
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• 유체기계공업학회 2005년도 연구개발 발표회 논문집
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• pp.503-508
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• 2005

An advanced Research Reactor (ARR) consists of an open-tank-type reactor assembly within a light water pool and generates thermal power of 20 MW. The thermal power is including a fission heat in the core, a fuel generated heat temporary stored in the pool, a circulating pumps generated heat and a neutron reflecting heat in the reflector vessel of the reactor. In order to remove the heat load, the primary cooling system will be installed. In this study, the conceptual design of the primary cooling system has been carried out using a design methodology of HANARO within a permissible range of safety. As results, it has been established that the conceptual design of the primary cooling system including design requirements, performance requirements, design restrictions, system descriptions and system operation to maintain the system functions.

• Nuclear Engineering and Technology
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• 제38권1호
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• pp.1-10
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• 2006

The interaction and mixing of high-temperature melt and water is the important technical issue in the safety assessment of water-cooled reactors to achieve ultimate core coolability. For specific advanced light water reactor (ALWR) designs, deliberate mixing of the core-melt and water is being considered as a mitigative measure, to assure ex-vessel core coolability. The paper provides the background of past experiments as well as key fundamentals that are needed for melt-water interfacial transport phenomena, thus enabling the development of innovative safety technologies for advanced LWRs that will assure ex-vessel core coolability.

• Nuclear Engineering and Technology
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• 제49권3호
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• pp.621-631
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• 2017

Many developing countries considering the introduction of nuclear power find that large-scale reactor plants in the range of 1,000 MWe to 1,600 MWe are not grid appropriate for their current circumstance. By contrast, small modular reactors are generally too small to make significant contributions toward rapidly growing electricity demand and to date have not been demonstrated. This paper proposes a radically simplified re-design for the nuclear steam cycle for a medium-sized reactor plant in the range of 600 MWe. Historically, balance of plant designs for units of this size have emphasized reliability and efficiency. It will be demonstrated here that advances over the past 50 years in component design, materials, and fabrication techniques allow both of these goals to be met with a less complex design. A disciplined approach to reduce component count will result in substantial benefits in the life cycle cost of the units. Specifically, fabrication, transportation, construction, operations, and maintenance costs and expenses can all see significant reductions. In addition, the design described here can also be expected to significantly reduce both construction duration and operational requirements for maintenance and inspections.