• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 춘계학술발표회논문집(2)
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• pp.271-276
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• 1996

Coolant flow distribution to the assemblies and core coolant/component temperatures should be determined in LMR core steady state thermal-hydraulic performance analysis. Sodium flow is distributed to core assemblies with the overall goal of equalizing the peak cladding midwall temperatures for the peak temperature pin of each pin bundle, thus pin cladding damage accrual and pin reliability. The flow orificing analysis for conceptual design will be performed with Excel spreadsheet program ORFCE which was set up and tested, using the calibration factors based on available analyses data. For the verification of this program, flow orificing calculation for the MDP 840MWth core was performed. The calculational results are satisfactory compared to those of CRIEPI calculation.

• Tribology and Lubricants
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• 제17권6호
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• pp.453-458
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• 2001

The fretting wear of a nuclear fuel rod it a dangerous phenomenon. In this study, TiN coating was used to reduce the fretting wear of Zircaloy-4 tube, a nuclear fuel rod cladding material. TiN coating is probably one of the molt frequently and successfully used PVD coatings for the mitigation of fretting wear. The fretting tester was designed and manufactured for this experiment. The number of cycles, slip amplitude and normal load were selected as main factors of fretting wear. The results of this research showed that wear volume was improved 1.3∼3.2 times with TiN coating. The worn surfaces were observed by SEM. Wear mechanism at lower slip amplitude was the brittle cracks and rupture of TiN coating. However, adhesive and abrasive wear were mainly observed on most surfaces at higher slip amplitude.

• Nuclear Engineering and Technology
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• 제43권6호
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• pp.499-508
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• 2011

The paper briefs a fuel performance code, COSMOS, which can be utilized for an analysis of the thermal behavior and fission gas release of fuel, up to a high burnup. Of particular concern are the models for the fuel thermal conductivity, the fission gas release, and the cladding corrosion and creep in $UO_2$ fuel. In addition, the code was developed so as to consider the inhomogeneity of MOX fuel, which requires restructuring the thermal conductivity and fission gas release models. These improvements enhanced COSMOS's precision for predicting the in-pile behavior of MOX fuel. The COSMOS code also extends its applicability to the instrumented fuel test in a research reactor. The various in-pile test results were analyzed and compared with the code's prediction. The database consists of the $UO_2$ irradiation test up to an ultra-high burnup, power ramp test of MOX fuel, and instrumented MOX fuel test in a research reactor after base irradiation in a commercial reactor. The comparisons demonstrated that the COSMOS code predicted the in-pile behaviors well, such as the fuel temperature, rod internal pressure, fission gas release, and cladding properties of MOX and $UO_2$ fuel. This sufficient accuracy reveals that the COSMOS can be utilized by both fuel vendors for fuel design, and license organizations for an understanding of fuel in-pile behaviors.

• Nuclear Engineering and Technology
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• 제52권3호
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• pp.597-609
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• 2020

Accident Tolerant Fuels have been widely studied since the Fukushima-Daiichi accident in 2011 as one of the options on how to further enhance the safety of nuclear power plants. Deposition of protective coatings on nuclear fuel claddings has been considered as a near-term concept that will reduce the high-temperature oxidation rate and enhance accidental tolerance of the cladding while providing additional benefits during normal operation and transients. This study focuses on experimental testing of Zr-based alloys coated with Cr-based coatings using Physical Vapour Deposition. The results of long-term corrosion tests, as well as tests simulating postulated accidents, are presented. Zr-1%Nb alloy used as nuclear fuel cladding serves as a substrate and Cr, CrN, Cr_xN_y layers are deposited by unbalanced magnetron sputtering and reactive magnetron sputtering. The deposition procedures are optimized in order to improve coating properties. Coated as well as reference uncoated samples were experimentally tested. The presented results include standard long-term corrosion tests at 360℃ in WWER water chemistry, burst (creep) tests and mainly single and double-sided high-temperature steam oxidation tests between 1000 and 1400℃ related to postulated Loss-of-coolant accident and Design extension conditions. Coated and reference samples were characterized pre- and post-testing using mechanical testing (microhardness, ring compression test), Thermal Evolved Gas Analysis analysis (hydrogen, oxygen concentration), optical microscopy, scanning electron microscopy (EDS, WDS, EBSD) and X-ray diffraction.

• Nuclear Engineering and Technology
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• 제28권2호
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• pp.150-161
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• 1996

연구로용 우라늄-실리사이드 분산 핵연료에서의 응력 및 변형율 분포를 계산할 수 있는 변형모델을 개발하였다. 이 변형모델은 탄소성이론 및 지수법칙 크리프이론을 기초로 한 것이며, 또한 등방 핵연료팽윤 및 열팽창을 가정하였다. 개발된 모델을 HANARO 및 카나다의 NRU 핵연료에 적용하여 본 결과 핵연료의 변형을 성공적으로 계산하는 것으로 판단되었다. 계산결과에 따르면, 연구로용 우라늄-실리사이드 분산핵연료가 연소할 때 핵연료심에서 가장 중요한 변형기구는 팽윤이며, 피복관에서 가장 중요한 변형기구는 크리프이다. 또한, 피복관에서 원주방향 최대응력은 항상 5 MPa 이하로서 항복응력보다 훨씬 낮게 유지되었다. 여기서 고려한 두 핵연료설계에 대해서 전 연소도 범위에서 핵연료봉의 부피변화는 10% 이하로 예측되었다.

• Nuclear Engineering and Technology
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• 제50권3호
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• pp.425-431
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• 2018

Different methods of axial and tangential testing and various sample geometries were investigated, and new test geometries were designed to determine the ultimate tensile strength of zirconium cladding tubes. The finite element method was used to model the tensile tests, and the results of the simulations were evaluated. Axial and tangential tensile tests were performed on as-received and machined fuel cladding tube samples of both E110 and E110G Russian zirconium alloys at room temperature to compare their ultimate tensile strengths and the different sample preparation methods.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2001년도 춘계학술발표회요약집
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• pp.274.1-274
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• 2001

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2003년도 춘계학술발표대회 요약집
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• pp.262.2-262.2
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• 2003

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1998년도 추계학술발표회요약집
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• pp.252-252
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• 1998

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1998년도 추계학술발표회요약집
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• pp.256-256
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• 1998