• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2018년도 추계학술논문요약집
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• pp.123-124
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• 2018

• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2018년도 추계학술논문요약집
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• pp.125-126
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• 2018

• 한국재료학회지
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• 제12권4호
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• pp.259-263
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• 2002

Three different ribbons of amorphous $Til_{1-x}Be_x$ alloys such as $Ti_{0.59}Be_{0.41},\;Ti_{0.61}Be_{0.39}\;and\;Ti_{0.63}Be_{0.37}$ were made by melt-spinning method to be used as brazing filler metals for joining Zircaloy-4 nuclear fuel cladding tubes, and their crystallization behavior as well as microstructure of the brazed zone were examined. The crystallization behavior was investigated in teams of thermal stability, crystallization temperature and activation energy. The crystallization of the $Ti_{1-x}Be_x$ alloys proceeded in two steps by the formation of ${\alpha}$-Ti at a lower temperature and of TiBe at a higher temperature. The crystallization temperature and activation energy of $Ti_{1-x}Be_x$ alloys were higher and larger than those of $Zr_{1-x}Be_x$ alloys and PVD Be. Those resulted thinner joining layer with $Ti_{1-x}Be_x$ alloys, which kept sound thickness of Zircaloy-4 nuclear fuel cladding tubes after brazing. But in the brazed zones made by $Ti_{1-x}Be_x$ filler metals, a little solid-solution layers composed of Zr and Ti were formed toward the Zr cladding tube and Zr was detected in the brazed zones. Microstructure of brazed zone was changed from globular to dentrite with decreasing Be content in the $Ti_{1-x}Be_x$ filler metal.

• Nuclear Engineering and Technology
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• 제43권2호
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• pp.149-158
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• 2011

Metal waste generated from the pyroprocessing of 10 MtU of spent fuel was classified by comparing the specific activity of a relevant radionuclide with the limit value of the specific activity specified in the Korean acceptance criteria for a lowand intermediate-level waste repository. A Korean Optimized Fuel Assembly design with a 17${\times}$17 array, an initial enrichment of 4.5 weight-percent, discharge burn-up of 55 GWD/MtU, and a 10-year cooling time was considered. Initially, the mass and volume of each structural component of the assembly were calculated in detail, and a source term analysis was subsequently performed using ORIGEN-S for these components. An activation cross-section library generated by the KENO-VI/ORIGEN-S module was utilized for top-end and bottom-end pieces. As a result, an Inconel grid plate, a SUS plenum spring, a SUS guide tube subpart, SUS top-end and bottom-end pieces, and an Inconel top-end leaf spring were determined to be unacceptable for the Gyeongju low- and intermediate-level waste repository, as these waste products exceeded the acceptance criteria. In contrast, a Zircaloy grid plate and guide tube can be placed in the Gyeongju repository. Non-contaminated Zircaloy cladding occupying 76% of the metal waste was found to have a lower level of specific activity than the limit value. However, Zircaloy cladding contaminated by fission products and actinides during the decladding process of pyroprocessing was revealed to have 52 and 2 times higher specific activity levels than the limit values for alpha and $^{90}Sr$, respectively. Finally, it was found that 88.7% of the metal waste from the 17${\times}$17 Korean Optimized Fuel Assembly design should be disposed of in a deep geological repository. Therefore, it can be summarized that separation technology with a higher decontamination factor for transuranics and strontium should be developed for the efficient management of metal waste resulting from pyroprocessing.

• 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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• 제55권7호
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• pp.2454-2465
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• 2023

Film boiling may lead to burnout of the heating element. Even though burnout does not occur, the heating element is subject to deformation because it is not sufficiently strong to withstand external loads. In particular, the ballooning and rupture of a tube under film boiling are important phenomena in the field of nuclear reactor safety. If the tube-type cladding of nuclear fuel ruptures owing to high internal pressure and thermal load, radioactive materials inside the cladding are released to the coolant. Therefore, predicting the ballooning and rupture is important. This study presents numerical simulations to predict the ballooning behavior and rupture time of a horizontal tube at high internal pressure under saturated film boiling. To do so, a multi-step coupled simulation of conjugated film boiling heat transfer and ballooning using creep model is adopted. The numerical methods and models are validated against experimental values. Two different nonuniform heat flux distributions and four different internal pressures are considered. The three-step simulation is enough to obtain a convergent result. However, the single-step simulation also successfully predicts the rupture time. This is because the film boiling heat transfer characteristics are slightly affected by the tube geometry related to creep ballooning.

• 한국윤활학회:학술대회논문집
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• 한국윤활학회 1996년도 제24회 춘계학술대회
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• pp.25-29
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• 1996

The fretting wear characteristics for Zircaloy-4 tube used as fuel rod in the nuclear power plant have been investigated. The fretting wear tester was designed and manufactured for this experiment. This study was focused on main factors of fretting wear, cycle, slip amplitude and normal load. The worn surfaces were observed by SEM.

• 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.

• 한국산학기술학회논문지
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• 제7권2호
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• pp.113-119
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• 2006

중수로형 핵연료 제조공정 중 연료봉 피복관에 간격체와 지지체 등의 부착물이 브레이징으로 접합된다. 본 연구에서는 베릴륨을 물리 증착법(PVD)으로 접합될 부착물의 표면에 증착한 것과 비정질 용가재[$Zr_{1-x}Be_{x}(0.3{\le}x{\le}0.5)$]를 사용하여 브레이징된 접합부의 미세조직과 경도 등의 특성을 비교하고 브레이징 온도가 접합부에 미치는 영향 조사하였다. 비정질 용가재에 의한 접합층의 두께는 PVD-Be의 경우와 비교하여 더 얇았고, Be 함량이 감소할수록 접합층의 두께는 감소하였으며 모재의 침식은 거의 없었다. PVD-Be의 경우 공정 반응, 액상 출현, 모세관 현상과 확산으로 브레이징 되나 비정질 합금은 용가재 만이 용융되어 액상 접합되는 것으로 사료된다. PVD-Be 접합부의 미세조직은 계면에서 수지상이 형성되어 내부로 성장하나, 비정질 합금에 의한 접합부는 석출된 제2상들이 구상으로 구성되며 브레이징 온도가 증가할수록 구상은 더욱 커졌다. 비정질 합금 접합부의 경도는 Be 함량이 감소할수록 경도는 증가하였다. 본 연구에 사용된 비정질 합금 중 $Zr_{0.7}Be_{0.3}$ 합금은 접합부에서 Be의 모재로의 확산이 적어 부드러운 계면과 모재의 침식이 없었고 높은 경도 때문에 핵연료 피복재 접합에 가장 적합한 용가재로 사료된다.

• Nuclear Engineering and Technology
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• 제17권3호
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• pp.211-215
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• 1985

온도 및 연신율 변촤가 Zircaloy-4의 요드 응력부식 거동에 미치는 영향을 30$0^{\circ}C$에서 일정 하중법과 300, 350, 40$0^{\circ}C$에서 일정 연신율법으로 ($10^{-5}$sec~$10^{-6}$ sec) 3.34mg $I_2$/㎤의 요드분위기에서 연구하였다. 요드 응력부식균열에 대한 저항성은 온도가 상승하거나 변형속도가 감소하면 감소했고 파손 시간과 응력과의 관계는1/tｆ∝exp (0.3$\sigma$/$\sigma$uTs-31.5)로 표시할 수 있었다. 30$0^{\circ}C$에서 요드 응력 부식 균열에 대한 저항성을 불활성 분위기에서의 파손에너지에 대한 요드분위기에서의 파손 에너지의 비율로 표시할 때 변형속도가 7.6$\times$$10^{-6}$ sec 부근에서 저항성이 가장 낮았고 온도가 35$0^{\circ}C$, 40$0^{\circ}C$ 로 증가함에 따라 보다 높은 변형속도에서 최저 저항성을 나타내는 경향을 보였다. 요드 응력부식 균열의 파단면에서 준-벽계 파면(quasi-cleavage fracture)을 관찰했다. 전술한 결과에 의하면 Zircaloy-4의 요드 응력부식균열의 기구에 있어서 보호 피막파손단계 (film rupture step)가 중요한 과정으로 추정된다.