• 분석과학
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• 제12권6호
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• pp.490-497
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• 1999

불활성기체용융-열전도도 측정법의 수소분석기를 개조하여 글로브박스에 설치했으며 조사핵연료피복관의 수소분석에 사용했다. Zr과 Ti 매질의 수소표준물질로 용융조제인 주석을 사용하지 않고도 수소함량 $3{\mu}g$까지 분석가능하였다. 시료의 크기가 작을수록 수소 회수율이 높았으며 지르칼로이 시료의 수소분석시 Ti 매질의 표준물질을 사용할 수 있음을 확인하였다. 수소분석에 사용한 실제 조사핵연료피복관의 평균 방사능은 10 mR/hr였으며 평균수소농도는 130 ppm이었다.

• 한국윤활학회:학술대회논문집
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• 한국윤활학회 2001년도 제34회 추계학술대회 개최
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• pp.19-26
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• 2001

The wear characteristic of Zircaloy-4 tube, which is used for a cladding of light water reactor fuel rod, is investigated experimentally. The experiment is conducted with contacting the crossed tube specimens in air as well as in water at room temperature with various combination of contact normal force and sliding distance of reciprocating motion. The contour and the volume of each wear are examined to study the effect of contact condition and environment on wear. As a result, it is found that the wear volume in the water environment is larger than that in the air for all the contact (i.e., force and sliding distance) conditions. However, the wear depth is greater in air than in water if the contact normal force and the sliding distance are larger. These are explained by the ease of detachment of wear particles from the contact surface. On the other hand, workrate model is applied with the contact shear force range measured by our wear tester. Investigated is the correlation between the workrate and the wear volume increase rate of the present experiment. The parabolic curve is found to fit well for the present wear data.

• Nuclear Engineering and Technology
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• 제10권2호
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• pp.65-72
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• 1978

원자로 가동시, 정상상태에서 벗어나 갑작스럽게 출력이 바뀔 때 발생하는 응력의 집중과 핵 분열시 발생하는 요오드의 부식에 의해서 생기는 피복물질의 응력 부식파괴현상을 이해하기 위하여, 이번 실험에서는 지르칼로이-4(Zicaloy-4)관을 사용하여 요오드응력부식 실험을 원자로 안의 상태에 가깝도록 30$0^{\circ}C$의 상태아래서 행하였다. 요오드 농도에 따라서 지르칼로이-4, 관(Tube)의 응력부식에 한 파괴시간을 구했고, 응력부식을 일으킬 수 있는 임계요오드 농도 및 임계접선방향의 응력을 구하였다. 요오드에 의한 응력부식이 화학석인 반응이라기 보다는 기계적인 반응성격을 갖기 때문에 응력부식을 파괴 역학적인 관점에서 설명하고자 응력과 파괴시간을 함수관계로 다음과 같이 표시해 보았다. log t$_{F}$ =5.5- (3/2) log$_{c}$-4log$\sigma$ t$_{F}$ : 파괴시간(") c : 요오드농도(mg/㎤) $\sigma$ : 응력 ($10^4$psi).

• Nuclear Engineering and Technology
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• 제9권3호
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• pp.151-163
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• 1977

CANDU-PHWR형 원자로인 월성 1호기의 Zircaloy-4 피복 핵연료 설계치를 중심으로 Segel method에 의하며 FACOM 230 OS$_2$/VS 콤퓨터 시스템을 사용하여 핵연료비를 계산하였다. 아울러 핵연료 제조공장의 수덩, 가동을, 곧장시설 낑산규모 증대, 건설지 및 운전비기 변동, 이자율의 변화, 원광가격의 물가상승을, 기술개발인자 등이 핵연료비 계산에 미치는 효과에 패한 민감도를 분석하였다.

• Nuclear Engineering and Technology
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• 제52권8호
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• pp.1742-1748
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• 2020

TSSD, TSSP, and TSSP2 of hydrogen for optimized-Zirlo (Zirlo™) alloy were measured by DSC in the range of 53-457 wppm. Solvus curves of the TSSs are derived and proposed in this study. The results show that the temperature gap between TSSD and TSSP solvus lines of Zirlo™ are similar to those of other zirconium alloys, but another gap between the TSSD and TSSP2 line differs significantly. In particular, the TSSP2 solvus line becomes closer to the TSSD solvus line than to TSSP unlike Zircaloy-4, so ΔT_TSSD-TSSP2 of Zirlo™ decreases with decreasing temperature. This implies that hydride reorientation can take place more significantly in Zirlo™ than in Zircaloy-4, and the limited temperature variation of 65 ℃ during the vacuum drying and the cooling-down process may not be sufficient to prevent the triggering of hydride reorientation in Zirlo™ cladding under long-term dry storage.

• Journal of Mechanical Science and Technology
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• 제15권9호
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• pp.1274-1280
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• 2001

The fretting wear behavior of the contact between Zircaloy-4 tube and Inconel 600, which are used as the fuel rod cladding and grid, respectively, in PWR nuclear power plants was investigated in air. In the study, number of cycles, slip amplitude and normal load were selected as the main factors of fretting wear. The results indicated that wear increased with load, slip amplitude and number of cycles but was affected mainly by the slip amplitude. SEM micrographs revealed the characteristics of fretting wear features on the surface of the specimens such as stick, partial slip and gross slip which depended on the slip amplitude. It was found that fretting wear was caused by the crack generation along the stick-slip boundaries due to the accumulation of plastic flow at small slip amplitudes and by abrasive wear in the entire contact area at high slip amplitudes.

• Nuclear Engineering and Technology
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• 제37권4호
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• pp.309-316
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• 2005

The contents of a new book currently in preparation are described. The dearth of books in the field of nuclear materials has left both students in nuclear materials classes and professionals in the same field without a resource for the broad fundamentals of this important sub-discipline of nuclear engineering. The new book is devoted entirely to materials problems in the core of light-water reactors, from the pressure vessel into the fuel. Key topics deal with the $UO_2$ fuel, Zircaloy cladding, stainless steel, and of course, water. The restriction to LWR materials does not mean a short monograph; the enormous quantity of experimental and theoretical work over the past 50 years on these materials presents a challenge of culling the most important features and explaining them in the simplest quantitative fashion. Moreover, LWRs will probably be the sole instrument of the return of nuclear energy in electric power production for the next decade or so. By that time, a new book will be needed.

• Nuclear Engineering and Technology
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• 제53권7호
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• pp.2295-2303
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• 2021

To investigate wall-thinning impact on axial load resistance of Zircaloy-4 cladding rods after a LOCA transient, axial tensile samples have been machined on as-received tubes with reduced thicknesses between 370 and 580 ㎛. After high temperature oxidation under steam at 1200 ℃ with measured ECR ranging from 10 to 18% and water quenching, machined samples were axially loaded until fracture. These tests were modeled using a fracture mechanics approach developed in a previous study. Fracture stresses are rather well predicted. However, the slightly lower fracture stress observed for wall-thinned samples is not anticipated by this modeling approach. The results from this study confirm that characterizing the axial load resistance using semi-integral tests including the creep and burst phases was the best option to obtain accurate axial strengths describing accurately the influence of wall-thinning at burst region.

• Nuclear Engineering and Technology
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• 제53권1호
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• pp.178-187
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• 2021

Dual-cooled annular fuel allows a significant increase in power density while maintaining or improving safety margins. However, the dual-cooled design brings much higher Zircaloy charge in reactor core, which could cause a great threaten of hydrogen explosion during severe accidents. Hence, an innovative fuel combined dual-cooled annular geometry and SiC cladding was proposed for the first time in this study. Capabilities of fuel design and behavior simulation were developed for this new fuel by the upgrade of FROBA-ANNULAR code. Considering characteristics of both SiC cladding and dual-cooled annular geometry, the basic fuel design was proposed and preliminary proved to be feasible. After that, a design optimization study was conducted, and the optimal values of as-fabricated plenum pressure and gas gap sizes were obtained. Finally, the performance simulation of the new fuel was carried out with the full consideration of realistic operation conditions. Results indicate that in addition to possessing advantages of both dual-cooled annular fuel and accident tolerant cladding at the same time, this innovative fuel could overcome the brittle failure issue of SiC induced by pellet-cladding interaction.

• 한국표면공학회:학술대회논문집
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• 한국표면공학회 2000년도 춘계학술발표회 초록집
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• pp.5-5
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• 2000

In the severe accident case like LOCA, Zircaloy(Zry) claddings are oxidized not only in high temperature but also in high pressures. It is a concem whether the safety of high bum up fuels can be maintained during severe accident. The effects of steam pressure on Zry-4 oxidation, and the effect of prc-existing oxide layer on the cladding in the high temperature-high pressure oxidation of Ziy-4 were investigated. The experimental temperature range was $700-900^{\circ}C$, and the pressures were between 0.1 and l5.0MPa. Partial pressure of steam tumed out to be the important one rather than total gas pressure. The higher the steam pressure was applied, the thicker the oxide became. nle effect of st,earn pressure on the oxidation of claddings with preexisting oxide was about 40-60% less effective than that of pickled cladding. Aocelerated oxidation in highpressure slean1 seems to be originated from the formation of microcracks produced during the transformation of tetragonal zirconia to monoclinic phase. Steam pressure seems to affect the stability of tetragonal phase.