• 대한기계학회:학술대회논문집
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• 대한기계학회 2001년도 춘계학술대회논문집A
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• pp.489-494
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• 2001

The fuel of light water reactor used far several years at high temperature and pressure, so it needs to clad with high corrosion resistance material. The cladding materials need low absorption of a neutron and high corrosion resistance. Cladding materials used Zircaloy-2 in Boiling Water Reactor, Zircaloy-4 in Pressurized Water Reactor and Zirlo has good for long term corrosion. If fracture of cladding tube occured during operation, it caused disaster. So it is needed to estimate of integrity fur cladding materials. In this paper, tension characteristics of cladding materials are investigate which is basic research far fracture characteristic. Also analysis of residual stress effect between tube type(original type) specimen and flattened type specimen.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2004년도 추계학술발표회 발표논문집
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• pp.931-932
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• 2004

• 한국재료학회지
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• 제29권8호
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• pp.483-490
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• 2019

Currently, the Korean nuclear industry uses ZIRLO as material for nuclear fuel cladding(zirconium alloy). KEPCO Nuclear Fuel is in the process of developing a HANA alloy to enable domestic production of cladding. Cladding manufacture involves multistage heat treatments and pickling processes, the latter of which is vital for the removal of defects and impurities on the cladding surface. SMUT that forms on the cladding surface during such pickling process is a source of surface defects during heat treatment and post-treatment processes if not removed. This study analyzes ZIRLO, HANA-4, and HANA-6 alloy claddings to extensively study the SEM/EDS, XRD, and particle size characteristics of SMUT, which are second phase particles that are formed on the cladding surface during pickling processes. Using the analysis results, this study observes SMUT formation characteristics according to Nb concentration in Zr alloys during the washing process following the pickling process. In addition, this study observes SMUT removal characteristics on cladding surfaces according to concentrations of nitric acid and hydrofluoric acid in the acid solution.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2008년도 추계학술대회A
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• pp.1377-1380
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• 2008

An irradiation test of a nuclear fuel rod having external and internal tube structure was planned for a performance. To establish fabrication process satisfying the requirements of irradiation test, micro-TIG welding system for fuel rods was developed, and preliminary welding experiments for optimizing process conditions of fuel rod was performed. Fuel rods with 15.9mm diameter and 0.57mm wall thickness of cladding tubes and end caps have been used and optimum conditions of endcap welding have been selected. In this experiment, the qualification test was performed by tensile tests, helium leak inspections, and metallography examinations to qualify the endcap welding procedure. The soundness of the welds quality of a dual cooled fuel rods has been confirmed by mechanical tests and microstructural examinations.

• 방사성폐기물학회지
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• 제20권2호
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• pp.161-170
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• 2022

The effect of oxidation time on the characteristics and mechanical properties of spent nuclear fuel cladding was investigated using Raman spectroscopy, tube rupture test, and tensile test. As oxidation time increased, the Raman peak associated with the tetragonal zirconium oxide phase diminished and merged with the Raman peak associated with the monoclinic zirconium oxide phase near 333 cm^-1. Additionally, the other tetragonal zirconium oxide phase peak at 380 cm^-1 decreased after 100 d of oxidation, whereas the zirconium monoclinic oxide peak became the dominant peak. The oxidation time had no effect on the tube rupture pressure of the oxidized zirconium alloy tube. However, the yield and tensile stresses of the oxidized nuclear fuel cladding tube decreased after 100 d of oxidation. The results of the scanning electron microscopy and transmission electron microscopy were represented with the in-situ Raman analysis result for the oxide characteristics generated on the cladding of spent nuclear fuel.

• 한국윤활학회:학술대회논문집
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• 한국윤활학회 2002년도 제35회 춘계학술대회
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• pp.88-95
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• 2002

Fretting of fuel rod cladding material, Zircaloy-4 Tube, in PWR nuclear power plants must be reduced and avoided. Nowadays the introduction of surface treatments or coatings is expected to bean ideal solution to fretting damage since fretting is closely related to wear, corrosion and fatigue. Therefore, in this study the fretting wear experiment was peformed using TiAIN coated Zircaloy-4 tube as the fuel rod cladding and uncoated Zircaioy-4 tube as one of grids, especially concentrating on the sliding component. Fretting wear resistance of TiAIN coated Zircaloy-4 tubes was improved compared with that of TiN coated tubes and uncoated tubes and the fretting wear mechanisms were delamination and plastic flow following by brittle fracture at lower slip amplitude but severe oxidation and spallation of oxidative layer at higher slip amplitude.

• 한국음향학회:학술대회논문집
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• 한국음향학회 1998년도 학술발표대회 논문집 제17권 2호
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• pp.247-250
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• 1998

The scattering of plane ultrasonic waves by the nuclear fuel pin of liquid metal reactor in sodium is studied. According to the internal composition in the cladding tube, the fuel pin has three cross sections, i.e. helium gas plenum, sodium-filled section, and fuel insertion section. The scattering spectra for each section of the fuel pin are different. The circumnavigating ultrasonic waves of each section are analyzed by the resonance scattering method. The whispering gallery wave modes are generated in the sodium-filled plenum section and the fuel rod insertion section with a sodium-gap. The circumferential wave modes are propagated in the cladding tube of the helium gas plenum section. The annular gap between the cladding tube and metal uranium pellet rod affects the scattering spectra. The different propagation characteristics can be utilized for the nondestructive method of detecting the unbonded area and measuring the level of the sodium-filled section of the fuel pin.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2001년도 춘계학술대회논문집A
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• pp.256-261
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• 2001

Corrosion and tensile properties of zircaloy-4 cladding tube having a laser welding part in elevated temperature are studied to present the criterion of quality evaluation in nuclear reactor and to found the scientific basis of SCC, with laser welding method using by coupling up cladding tube to end cap. In the result of tensile test($400^{\circ}C$), the fracture is not happened in the welding part but base metal and the result of corrosion test($400^{\circ}C$ 1500psi steam), corrosion rate of the molten zone and PMZ is a little higher than the other zone.

• 대한용접접합학회:학술대회논문집
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• 대한용접접합학회 2001년도 추계학술발표대회 개요집
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• pp.65-68
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• 2001

Corrosion and tensile properties of zircaloy-4 cladding tube having a laser welding part in elevated temperature are studied to present the criterion of quality evaluation in nuclear reactor and to found the scientific basis of SCC, with laser welding method using by coupling up cladding tube to end cap. In the result of tensile test(40$0^{\circ}C$), the fracture is not happened in the welding part but base metal and the result of corrosion test(40$0^{\circ}C$ 1500psi steam), corrosion rate of the molten zone and PMZ is a little higher than the other zone

• 한국재료학회지
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• 제4권4호
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• pp.445-452
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• 1994

가압 경수로 핵연료의 중성자 조사 조건에서 Zircaloy피복관의 3축방향으로의 변동거동은 집합도 계수에 따른 크립 이방성고 조사성장 이방성을 통하여 분석될 수 있다. 이러한 크립과 조사성장의 이방성이 Zircaloy피복관의 각 축방향 변형율에 미치는 영향을 평가할 수 있는 방법론이 제시되었다. 연소 후 측정된 KOFA Zircaloy-4피복관의 변형율과 핵연료 성능 분석 코드의예측치를 토대로 하여 각 축방향 변형율을 계산한 결과 KOFA Aircaloy-4 피복관의 원주방향 변형은 크립에 의해 주로 일어난 반면, 피복관의 길이방향 변형은 조사성장에 의하여 일어났으나 낮은 조사량에서는 크립의 영향도 상당히 큰것으로 나타났다.