• 한국표면공학회지
• /
• 제47권1호
• /
• pp.7-12
• /
• 2014

The characteristics of oxidation for the Zry-4 was measured in the $800^{\circ}C$ and high steam pressure (50 bar, 75 bar, 100 bar) conditions, using an apparatus for high pressure steam oxidation. The effect of accelerated oxidation by high-pressure steam was increased more than 60% in hydrogen-charged cladding than normal cladding. This difference between hydrogen charged claddings and normal claddings tends to be larger as the higher pressure. The accelerated oxidation effect of hydrogen charging cladding is regarded as the hydrogen on the metal layer affects the formation of the protective oxide layer. The creation of the sound monoclinic phase in Zry-4 oxidation influences reinforcement of corrosion-resistance of the oxide layer. The oxidation is estimated to be accelerated due to the creation of equiaxial type oxide film with lower corrosion resistance than that of columnar type oxide film. When tetragonal oxide film transformed into the monoclinic oxide film, surface energy of the new monoclinic phase reduced by hydrogen in the metal layer.

• Corrosion Science and Technology
• /
• 제18권5호
• /
• pp.175-181
• /
• 2019

A failed spent fuel rod with 53,000 MWd/tU from a nuclear power plant was characterized, and the fission products and oxygen layer in the pellet-clad mechanical interaction region were observed using an EPMA (Electron Probe Micro-Analyzer). A sound fuel rod burned under similar conditions was used to compare and analyze, the results of the failed fuel rod. In the failed fuel rod, the oxide layer represented $10{\mu}m$ of the boundary of the cladding, and $35{\mu}m$ of the region outside the cladding. By comparison, in the sound fuel rod, the oxide layer was $8{\mu}m$, observed in the cladding boundary region. The cladding inner surface corrosion and the resulting fuel-cladding bonding were investigated using an EPMA. Zirconium existed in the bonding layer of the (U, Zr)O compound beyond the pellet cladding interaction gap of $20{\mu}m$, and composition of UZr2O3 was observed in the failed fuel rod. This paper presents the results of the EPMA examination of a spent fuel specimen, and a technique to analyze fission products in the pellet-clad mechanical interaction region.

• 한국압력기기공학회 논문집
• /
• 제19권2호
• /
• pp.109-116
• /
• 2023

This study presents a CRUD modelling to simulate the thermal resistance behavior of CRUD, deposited on the surface of a cladding tube of a fuel assembly. When heat produced from fuels transfers to a coolant through a cladding tube, the CRUD acting as an additional thermal resistance is expressed as two layers, i.e., a solid oxide layer and an imaginary fluid layer, which are added to the experimental tube's heat structure of the MARS-KS input data. The validation calculation for the experiments performed in UNIST-DISNY facility showed that the center and surface temperatures of the cladding tube increased as the porosity and the steam amount inside pores of the CRUD got higher. In addition, the temperature gradient in the imaginary fluid layer was calculated to be larger than that in the solid oxide part, indicating that the steam amount inside the layer acted more largely as thermal resistance. It was also evaluated through sensitivity calculations that the cladding tube temperature was more sensitive to the CRUD porosity and the steam amount in pores than to the inlet flow rate of the coolant.

• 한국원자력학회:학술대회논문집
• /
• 한국원자력학회 1995년도 추계학술발표회논문집(2)
• /
• pp.680-686
• /
• 1995

It has been known that water-side corrosion of fuel rods in nuclear reactor is accompanied with the loss of metallic wall thickness and pickup of hydrogen. This corrosion is one of the important limiting factors ill the operating life of fuel rods. In connection with the fuel cladding corrosion, a device to measure the water-side oxide layer thickness by means of the eddy-current method without destructing the fuel rod was developed by KAERI. The device was installed on the multi-function testing bench in the nondestructive test hot-cell and its calibration was carried out successfully for the standard rod attached with plastic thin films whose thicknesses are predetermined. It shows good precision within about 10% error. And a PWR fuel rod, one of the J-44 assembly discharged from Kori nuclear power plant Unit-2, has been selected for oxide layer thickness measurements. With the result of data analysis, it appeared that the oxide layer thicknesses of Zircaloy cladding vary with the length of the fuel rod, and their thicknesses were compared with those of the destructive test results to confirm the real thicknesses.

• 한국정밀공학회:학술대회논문집
• /
• 한국정밀공학회 2002년도 춘계학술대회 논문집
• /
• pp.844-847
• /
• 2002

This paper is to investigate the wear behavior of submerged-arc clad materials by the wear test with a ball-on-disk type wear testing machine in air. The specimens were clad with Stoody105 alloy wire on a carbon steel (SM45C) substrate by submerged-arc cladding process under different welding parameters. The wear behavior of the cladding through ball-en-disk test has been studied under the wear load from 5N to 16N and sliding speed from 8cm/s to 35cm/s. The weight of the specimen loss was measured. Scanning electron micrographs of the worn surface show a layer of oxide film formed on the worn surface. Oxidation wear mechanism controls the wear process. The spalling of the oxide is caused by the repeated rubbing fatigue mechanism.

• 한국정밀공학회지
• /
• 제20권1호
• /
• pp.179-186
• /
• 2003

This paper is to investigate the wear behavior of submerged-arc claddings by the wear test with a ball-on-disk type wear testing machine in air. The specimens were clad with Stoody105 alloy wire on a medium carbon steel (SM45C) substrate by submerged-arc cladding process under different welding parameters. The wear behavior of the cladding through ball-on-disk test has been studied under the wear load from 5 to 16 N and the sliding speed from 8 to 35 cm/s. The weight loss of the specimen was measured. Scanning electron micrographs of the worn surface show a layer of oxide film formed on the worn surface. Oxidation wear mechanism controls the wear process. The spatting of the oxide is caused by the repeated rubbing fatigue mechanism.

• Nuclear Engineering and Technology
• /
• 제41권2호
• /
• pp.163-170
• /
• 2009

In order to evaluate high-temperature oxidation behavior of the advanced alloy cladding under LOCA conditions, isothermal oxidation tests in steam were performed with cladding specimens prepared from high burnup PWR fuel rods that were irradiated up to 79 MWd/kg. Cladding materials were $M5^{(R)}$ and $ZIRLO^{TM}$, which are Nb-containing alloys. Ring-shaped specimens were isothermally oxidized in flowing steam at temperatures from 1173 to 1473 K for the duration between 120 and 4000s. Oxidation rates were evaluated from measured oxide layer thickness and weight gain. A protective effect of the preformed corrosion layer is seen for the shorter time range at the lower temperatures. The influence of pre-hydriding is not significant for the examined range. Alloy composition change generally has small influence on oxidation in the examined temperature range, though $M5^{(R)}$ shows an obviously smaller oxidation constant at 1273 K. Consequently, the oxidation rates of the high burnup $M5^{(R)}$ and $ZIRLO^{TM}$ cladding are comparable or lower than that of unirradiated Zircaloy-4 cladding.

• 한국재료학회지
• /
• 제14권7호
• /
• pp.462-469
• /
• 2004

The effect of Cu addition on the corrosion characteristics of Zr alloys that developed for nuclear fuel cladding in KAERI (Korea Atomic Energy Research Institute) was evaluated. The alloys having different element of Nb, Sn, Fe, Cr and Cu were manufactured and the corrosion tests of the alloys were performed in static autoclave at $360^{\circ}C$, distilled water condition. The alloys were also examined for their microstructures using the optical microscope and the TEM equipped with EDS and the oxide property was characterized by using X-ray diffraction. From the result of corrosion test more than 450 days, the corrosion rate of the Zr-based alloys was changed with alloying element such as Nb, Sn, Fe, Cr and especially affected by Cu addition. The corrosion resistance was increased with increasing the Cu content and the tetragonal $ZrO_2$ layer was more stabilized on the Cu-containing alloys.

• 대한전자공학회논문지SD
• /
• 제41권9호
• /
• pp.31-37
• /
• 2004

oxide 층의 위치와 두께 조절을 통하여 oxide-VCSEL이 단일모드로 동작하는 활성영역의 반경을 확장하는 방법을 Self-consistent effective-index 방법을 이용하여 제시하였다. 이렇게 활성영역이 넓어지면 고속, 고신뢰도, 고출력 동작에 유리한 단일모드 VCSEL을 만들 수 있게 된다. 고출력을 위하여 단일모드로 동작하는 활성영역을 확대하는 방법을 하면 다음과 같다. 첫째 oxide 층은 활성층에서 멀리 떨어진 곳에 위치시켜야 한다 둘째, oxide 층은 얇게 만들어야 한다. 셋째, oxide층을 node에 위치시켜야 한다. 그리고 고출력을 위하여 p-DBR 쌍의 수를 줄이는 것은 단일모드 동작조건을 변화시키지 않는다. 이 방법을 사용하면 단일모드로 동작하는 oxide-aperture 크기를 3m% 이상 키울 수 있다.

• Nuclear Engineering and Technology
• /
• 제26권4호
• /
• pp.526-535
• /
• 1994

KMRR(다목적 연구용원자로) 핵연료의 알루미늄 피복재의 부식거동을 평가하기 위해, 부식 예측치와 노내 부식 실측치의 비교를 통해 유도된 열속인자를 도입한 수정된 Griess 경험식을 유도하였다. KMRR 핵연료의 건전성이 유지되는 부식의 설계기준으로써는 산화층의 박리 방지가 보수적으로 설정되었으며, 산화층의 박리는 산화층에서의 온도차이가 114$^{\circ}C$ 이상에서 일어난다고 보수적으로 가정하였다. KMRR 핵연료의 출력이력을 첫 주기부터 평형주기까지 분석하여, 한계출력이력을 결정하였다. 한계출력이력을 가진 KMRR 핵연료의 부식량 예측계산 결과, 최대 산화층의 두께는 50$\mu\textrm{m}$ 이하였으며, 산화층 박리의 설계기준은 2배의 여유도를 가지고 만족하였다. 따라서, KMRR 핵연료는 피복재의 부식으로 인해 손상되지 않을 것으로 판단된다. 그러나, 수정된 Griess 부식경험식의 KMRR에의 적용 타당성은 KMRR 핵연료의 부식 감시를 통해 추가로 검증될 필요성이 있다.