• Nuclear Engineering and Technology
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• 제52권3호
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• pp.508-519
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• 2020

During transients or accidents, the reactor core is uncovered, and droplets entrained above the quench front collides with the uncovered fuel rod surface. Droplet impact cooling can reduce the peak cladding temperature. Besides zirconium-based cladding, versatile accidental tolerant fuel (ATF) claddings, including FeCrAl, have been proposed to increase the accident coping time. In order to investigate the effect of surface properties on droplet impact cooling of cladding surfaces, the droplet impact phenomena are photographed on the FeCrAl and zircaloy-4 (Zr-4) surfaces under different conditions. On the oxidized FeCrAl surface, the Leidenfrost phenomenon is not observed even when the surface temperature is as high as 550 ℃ with We > 30. Comparison of the impact behaviors observed on different materials shows that nucleate and transition boiling is more intensive on surfaces with larger thermal conductivity. The Leidenfrost point temperature (LPT) decreases with the solid thermal effusivity (${\sqrt{k{\rho}C_p}}$). However, the CHF temperature is relatively insensitive to the surface oxidation and Weber number. Droplet spreading diameter is analyzed quantitatively in the film boiling stage. Based on the energy balance a correlation is proposed for droplet maximum spreading factor. A mechanistic model is also developed for the LPT based on homogeneous nucleation theory.

• 한국산학기술학회논문지
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• 제9권2호
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• pp.276-280
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• 2008

본 연구에서 개발할 설계기술은 부가가치가 매우 높은 소량 다품종 건축용 내 외장재의 금속 패널 코너가공 법에 관한 것이다. 금속 금구류 및 건축용 내 외장재로 사용되는 2.5mm이상 되는 금속강판을 사용함에 있어서 가장 문제가 되는 직각코너링 접합부위에서 발생되는 내식성(수명성), 내후성, 및 미려성(디자인)을 향상시키므로 품질의 향상뿐만 아니라 소량 다품종의 생산성을 증대시킬 수 있는 국내 최초 금속패널 코너가공을 할 수 있는 벤딩 다이시스템(Bending Die System)을 설계하고자 한다.

• 한국산학기술학회논문지
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• 제9권6호
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• pp.1518-1522
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• 2008

본 연구에서 개발한 벤딩다이시스템(Bending Die System)제작기술은 부가가치가 매우 높은 소량 다품종 건축용 내 외장재의 금속 패널 코너가공법에 관한 것이다. 금속 금구류 및 건축용 내 외장재로 사용되는 2.5mm이상 되는 금속강판을 사용함에 있어서 가장 문제가 되는 직각코너링 접합부위에서 발생되는 내식성, 내후성, 및 디자인을 향상시키므로 품질의 향상뿐만 아니라 소량 다품종의 생산성을 증대시킬 수 있는 국내 최초 금속패널 코너가공을 할 수 있는 벤딩다이시스템을 설계하여 제작하였다. 이를 제작하여 경제적, 기술적으로 고부가가치를 창출하여 수출증대에 기여할 수 있다.

• Nuclear Engineering and Technology
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• 제48권2호
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• pp.572-594
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• 2016

As one of the Gen-IV nuclear energy systems, a sodium-cooled fast reactor (SFR) is being developed at the Korea Atomic Energy Research Institute. As a long-term national research project, advanced radiation resistant oxide dispersion strengthened steel (ARROS) is being developed as an in-core fuel cladding tube material for a SFR in the future. In this paper, the current status of ARROS development is reviewed and its future prospective is discussed.

• Nuclear Engineering and Technology
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• 제50권3호
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• pp.504-511
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• 2018

In Lithuania, all spent nuclear fuel (SNF) resulted from the operation of the Ignalina Nuclear Power Plant (INPP), which had two Russian Acronym for "Channelized Large Power Reactor"-type reactors. After the final shutdown, the total amount of SNF at the INPP was approximately 22,000 fuel assemblies. All these assemblies will be stored for about 50 years and disposed of after that. The decision to shut down and decommission both reactors in Lithuania before termination of design period raises a significant challenge for the treatment of accumulated SNF. Therefore, various techniques and technologies for SNF management were developed and justified for that specific case, and a set of special equipment was installed at the INPP, the effectiveness of which was demonstrated during its operation. This article presents unique techniques related to the management of SNF adopted and commissioned at the INPP after its operation shutdown, namely fuel rod cladding leak tightness control system and special equipment for collection of possible spillage during handling of SNF assembly in the hot cell. The operational experience and measurement results of fuel rod cladding leak tightness control system are presented.

• 한국광학회지
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• 제23권4호
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• pp.135-142
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• 2012

본 연구에서는 공동경로간섭계(common-path interferometer)에 기반한 회절위상현미경(diffraction phase microscopy)을 이용한 광섬유의 굴절률 프로파일(refractive index profile) 측정기술을 개발하였다. 투과형 회절격자를 이용하여 광섬유 시료를 통과한 빛으로부터 핀홀을 이용하여 영의 공간주파수 성분만을 갖는 기준광을 생성하고, 기준광을 다시 시료의 위상정보를 갖는 시료광과 간섭시키는 방법을 통해 시료의 위상정보를 가진 간섭무늬를 형성시켰다. 이렇게 얻어진 간섭 이미지로부터 수치적 처리과정을 거쳐 공간적 위상정보 곧, 위상 이미지를 획득하고 이 데이터를 역아벨변환(inverse Abel transform)을 통해 굴절률 프로파일로 변환할 수 있었다. 이때 클래딩과 광섬유 주변의 매질 사이의 굴절률차로 인해 발생하는 배경위상을 이론적으로 얻어진 함수형태에 맞춰 예측하고 이를 측정된 위상에서 제거하는 배경위상제거 방법을 개발하여 사용하였다. 이를 통해 광섬유 코어 부근의 위상정보만으로도 굴절률 프로파일을 성공적으로 이뤄질 수 있음이 입증되었다. 본 연구를 통하여 회절위상현미경 특유의 측정 안정성과 편의성을 가진 광섬유 굴절률 프로파일 측정장치를 개발하였고 광섬유 및 도파로의 굴절률 분포를 비파괴적으로 분석할 수 있어 광섬유 및 광섬유소자 개발에 활용될 수 있을 것으로 기대된다.

• 열처리공학회지
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• 제17권5호
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• pp.271-277
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• 2004

Zr-xNb alloys (x = 0.2, 0.8, 1.5 wt.%) were prepared to study the characteristics of the phase transformation in Zr-Nb system. The samples were heat treated at ${\beta}$-temperature ($1020^{\circ}C$) for 20 min and then cooled with different cooling rate. The microstructures of the specimens having the same compositions were changed with cooling rate and Nb content. The Widmanst$\ddot{a}$tten structure was observed on the furnace-cooled sample. The relationship between ${\alpha}$-Widmanst$\ddot{a}$tten and ${\beta}$-phase was the {0001}${\alpha}$//{110}${\beta}$, <11$\bar{2}$0>//<111>. The ${\beta}$-phase in Widmanst$\ddot{a}$tten structure of Zr-Nb alloys containing Nb more than solubility limit was identified as ${\beta}_{Zr}$ phase which was a stable phase at high temperature. In the water quenched samples, two kinds of martensite structures were observed depending on the Nb-concentration. The lath martensite was formed in Zr-0.2, 0.8 wt.% Nb alloys and the plate martensite having twins was formed in Zr-1.5 wt.% Nb alloy.

• Nuclear Engineering and Technology
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• 제20권3호
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• pp.170-178
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• 1988

중수로형 핵연료의 피복재로 사용되는 Zircaloy-4관의 결함검사를 위한 자동초음파 탐상 장치가 개발되었다. 이 장치에는 중심진동수가 14 MHz이고 대역폭이 11MHz인 집속 초음파 펄스를 사용한 수침 펄스-에코우 탐상기술과 특별히 고안된 시험수조 이송식 초음파주사 기술이 적용되었다 같은 크기와 방향을 갖는 관내외면 결함들을 같은 높이의 초음파 신호로 검출하기 위한 초음파 빔의 최적입사각은 26도이었다. Zircaloy-4피복관의 최대 허용 결함인, 깊이가 관두께의 10%인 0.04 mm이고, 길이가 0.76 mm인 축방향 및 길이가 0.38 mm인 원주방향 V형 인공결함들이 관내외면에 개재된 표준시험관을 사용하여 이 장치의 성능시험을 수행하였다. 그 결과 인공 표준시험관내의 모든 결함들을 매우 우수한 재현성을 갖고 분당 약 1m의 속도로 검출할 수 있었으며 이때의 신호 대 잡음 비는 축방향 결함에 대해서는 20 dB, 원주방향 결함에 대해서는 12 dB 이상이었다.

• Nuclear Engineering and Technology
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• 제48권5호
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• pp.1096-1108
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• 2016

Metal fuel is being developed for the prototype generation-IV sodium-cooled fast reactor (PGSFR) to be built by 2028. U-Zr fuel is a driver for the initial core of the PGSFR, and U-transuranics (TRU)-Zr fuel will gradually replace U-Zr fuel through its qualification in the PGSFR. Based on the vast worldwide experiences of U-Zr fuel, work on U-Zr fuel is focused on fuel design, fabrication of fuel components, and fuel verification tests. U-TRU-Zr fuel uses TRU recovered through pyroelectrochemical processing of spent PWR (pressurized water reactor) fuels, which contains highly radioactive minor actinides and chemically active lanthanide or rare earth elements as carryover impurities. An advanced fuel slug casting system, which can prevent vaporization of volatile elements through a control of the atmospheric pressure of the casting chamber and also deal with chemically active lanthanide elements using protective coatings in the casting crucible, was developed. Fuel cladding of the ferritic-martensitic steel FC92, which has higher mechanical strength at a high temperature than conventional HT9 cladding, was developed and fabricated, and is being irradiated in the fast reactor.

• Nuclear Engineering and Technology
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• 제54권8호
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• pp.3154-3165
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• 2022

During a loss-of-coolant accident (LOCA) in the pressurized water reactor (PWR), there is a possibility that high temperature and internal pressure of the fuel rods lead to ballooning of the cladding, which causes a partial blockage of flow area in a subchannel. Such flow blockage would influence the core coolant flow, thus affecting the core heat transfer during a reflooding phase and subsequent severe accident. However, most of the system analysis codes simulate the accident process based on the assumed channel blockage ratio, resulting in the fact that the simulation results are not consistent with the actual situation. This paper integrates the developed core Fuel Rod Thermal-Mechanical Behavior analysis (FRTMB) module into the self-developed severe accident analysis code ISAA. At the same time, the existing flow blockage model is improved to make it possible to simulate the change of flow distribution due to fuel rod deformation. Finally, the ISAA-FRTMB is used to simulate the QUENCH-LOCA-0 experiment to verify the correctness and effectiveness of the improved flow blockage model, and then the effect of clad ballooning on core heat transfer and subsequent parts of core degradation is analyzed.