• Proceedings of the Korean Society of Tribologists and Lubrication Engineers Conference
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• 2001.06a
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• pp.81-88
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• 2001

To study the influence of the shape of contacting bodies (especially the end profile) on slip regime, wear test is conducted in the case of the contact between tube and support. Two different end profiles of the support are used such as truncated wedge and rounded punch. During the test, 10, 30 and 50 N are applied as normal force and slip displacement varies between 10-200 $\mu\textrm{m}$. The tube and the support specimens are made of Zircaloy-4 and a specially designed wear tester is used. Tests are carried out in air at room temperature. Wear on the tube is examined by measuring microscope. Partial and gross slip regimes are classified from the observed wear shape. Surface roughness tester is also used to measure the wear depth and contour, from which wear volume is evaluated. The transition from partial to gross slip is also investigated by investigating the considerable increase of wear volume. From the result, the boundary between the partial and the gross slip is newly determined in the conventional fretting map for the present specific contact configuration. Since the transition is related with the amount of energy dissipation from the contact surface so is wear, it is regarded that wear can be restrained by designing a proper shape of support.

• Proceedings of the Korean Nuclear Society Conference
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• 1995.05a
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• pp.805-805
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• 1995

핵연료저장시설의 화재 등 극단적인 사고조건하에서 $UO_2$ 소결펠렛의 습식산화와 건식산화에 대한 연구를 수행하였다. 손상된 지르칼로이 피복관 속의 $UO_2$ 소결펠렛을 산성분위기의 습윤조건하에서 산화시킬 때의 $UO_2$ 펠렛의 산화속도는 IDR(mg/$\textrm{cm}^2$.min) = 1.55 [H$^{+}$]$^{1.21}$ 로 나타났다. 또한 습윤조건하에서 $UO_2$ 분말에 알카리 및 알카리 토금속 산화물, 그리고 백금족 및 회토류 산화물 등과 같은 불순물들이 존재할 때의 산화속도를 조사하였으며 이들에 대한 영향도 관찰하였다. 핵연료저장시설의 가상화재를 바탕으로 한 400~$700^{\circ}C$의 온도범위에서, 피복관이 씌워진 $UO_2$ 소결펠렛의 건식산화반응을 조사한 바 $UO_2$ 소결펠렛은 산화초기에 U$_4$O$_{9}$ 또는 U$_3$O$_{7}$ 등의 중간상 형성에 따른 3-4%의 부피축소에 의해 결정립계 균열이 일어나고, $600^{\circ}C$ 이하에서는 온도증가에 따라 중간상에서 U$_3$O$_{8}$ 상으로의 상변화에 의한 부피팽창으로 피복관의 변형과 함께 산화속도의 가속을 발견할 수 있었고, $600^{\circ}C$ 이상에서는 핵연료소자의 소성변형으로 인한 산화속도의 지연을 발견할 수 있었다. 또한 $UO_2$ 펠렛의 건식산화거동은 기체-고체 반응시의 전형적인 형태인 shrinking core model에 잘 적용될 것으로 판단되었다.

• Proceedings of the Materials Research Society of Korea Conference
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• 2003.11a
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• pp.213-213
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• 2003

고온 및 고압의 가혹한 방사선 분위기에서 사용되는 핵연료 피복관은 중성자 조사 및 수소화합물의 생성 등으로 인하여 기계적 성질이 저하된다. 따라서 조사된 핵연료 피복관의 손상기준 확립과 안전성 해석을 위해서는 연성 및 강도 등 기계적 특성을 정확히 이해하여야 할 필요가 있다. 핵연료 피복관의 종 및 횡 방향 인장특성 평가를 위하여 개발된 기존의 다양한 시험법들을 비교하고, 핫셀시험에 적합한 인장시험법을 개발하였다. 피복관의 종방향 인장시편은 튜브시편 또는 게이지부 내에서 균일한 변형률 분포를 얻도록 설계된 도그본 튜브시편(그림 1)을 사용한다. 피복관의 횡방향 인장시험에 사용되는 링시편(그림 2)은 게이지부 내에서 균일한 단축 원환변형율 분포 또는 평면변형율 조건을 나타내도록 설계한다. 연소 또는 조사된 피복관으로부터 시편을 제작하기 위해서는 핫셀 내에서 작업 이 가능한 방전가공기(그림 3)를 사용한다. 피복관의 종방향 인장시험용그립(grip)은 핀-부하형이며, 횡방향 인장시험의 경우는 시험 동안 시편의 곡률이 일정하게 유지 되도록 그립의 형상 및 치수를 결정한다(그림 4). 피복관의 종 및 횡방향 강도와 변형 등 기계적 특성을 평가하기 위한 응력-변형율 곡선은 시험기의 복합 강성(K)을 고려하여 결정한다. 이상과 같이 검토된 인장시험법은 피복관의 안전성 해석(safety analysis)과 관련 규정(regulatory)에서 사용되는 피복관 손상기준(fuel damage criteria)의 개선에 필수적인 자료를 제공한다.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.37 no.10
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• pp.1239-1249
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• 2013

In this study we establish a process to predict hardening behavior considering the Bauschinger effect for zircaloy-4 sheets. When a metal is compressed after tension in forming, the yield strength decreases. For this reason, the Bauschinger effect should be considered in FE simulations of spring-back. We suggested a suitable specimen size and a method for determining the optimum tightening torque for simple shear tests. Shear stress-strain curves are obtained for five materials. We developed a method to convert the shear load-displacement curve to the effective stress-strain curve with FEA. We simulated the simple shear forward/reverse test using the combined isotropic/kinematic hardening model. We also investigated the change of the load-displacement curve by varying the hardening coefficients. We determined the hardening coefficients so that they follow the hardening behavior of zircaloy-4 in experiments.

• Korean Journal of Materials Research
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• v.12 no.4
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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.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.37 no.1
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• pp.67-75
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• 2013

The flow stress of zircaloy-4 used in the spacer grid supporting a nuclear fuel rod was determined by the Johnson-Cook model, and model parameters were determined using reverse engineering. Parameters such as A, B, n and $\dot{\varepsilon}_0$ were determined by the tensile test result. To obtain the parameters C and m, a slot milling test and numerical simulation were performed. The objective functions were defined as the difference between the experimental and the simulation results, and then, the parameters were determined by minimizing the objective function. To verify the validity of the determined parameters, cross-verification for each case was conducted through a shearing test and simulation. The results tend to show agreement with the experimental results, such as the features of sheared edges and maximum punch force, with the correlation coefficients exceeding at least 0.97.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.10 no.4
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• pp.273-280
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• 2012

The electrochemical behaviors of 10 g-scale fresh and oxidized Zircaloy-4 cladding hulls were examined in $500^{\circ}C$ LiCl-KCl molten salts to confirm the feasibility of the electrorefining process for the treatment of hull wastes. In the results of measuring the potential-current response using a stainless steel basket filled with oxidized Zircaloy-4 hull specimens, the oxidation peak of Zr appears to be at -0.7 to -0.8 V vs. Ag/AgCl, which is similar to that of fresh Zircaloy-4 hulls, while the oxidation current is found to be much smaller than that of fresh Zircaloy-4 hulls. These results are congruent with the outcome of current-time curves at -0.78 V and of measuring the change in the average weight and thickness after the electrochemical dissolution process. Although the oxide layer on the surface affects the uniformity and rate of dissolution by decreasing the conductivity of Zircaloy-4 hulls, electrochemical dissolution is considered to occur owing to the defect of the surface and phase properties of the Zr oxide layer.

• Nuclear Engineering and Technology
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• v.23 no.1
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• pp.1-11
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• 1991

The objective of this study is to investigate the effect of weld line on the mechanical strengths and the process of weld line elimination in the Zircaloy-4 resistance upset welding for the fabrication of heavy water reactor fuel rods. The weld current and the amount of upset increased linearly with the main heat, in which two relations between them were derived. It was found that the threshold to obtain sound weld was 50% of main heat in terms of weld upset size, mechanical strengths and weld line elimination. The weld microstructure of resistance upset welds of Zircaloy-4 comprsied basketweave, Widmanstatten and martensite respectively by changing the main heats. Dimples on uniaxially fractured surface at weld line in the Zr-4 welds were larger and deeper compared with those on biaxially fractured surface. It was also found that the process of the weld line elimination in the resistance upset weld of Zircaloy-4 could be divided into three stages in terms of the presence of many pores, their shrinkage and elimination, and the shrinkage of the original weld interface with increasing weld currents.

• Nuclear Engineering and Technology
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• v.17 no.3
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• pp.211-215
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• 1985

The effects of temperature and strain rate on the I-SCC behaviors of Zircaloy-4 were investigated by constant load test at 30$0^{\circ}C$ and constant elongation rate test at 300, 350 and 40$0^{\circ}C$ in 3.34mg $I_2$/㎤. The results showed that I-SCC susceptibility increased as the strain rate decreased or the temperature increased. The empirical relation between the stress and the time to failure at 30$0^{\circ}C$ was given by 1/ $t_{f}$∝exp (0.3$\sigma$/$\sigma$$_{UTS}$-31.5) When the I-SCC susceptibility was expressed by the ratio of fracture energy in iodine atmosphere to that in the inert atmosphere, severe I-SCC susceptibility was found near 7.6$\times$10$^{-6}$ sec at 30$0^{\circ}C$ and the maximum point of I-SCC susceptibility tended to shift to the higher strain rate with increasing the temperature. The quasi-cleavage fracture was observed in I-SCC fracture surface. From these results, it was certain that the film repture step was involved as an important process in the I-SCC mechanism of Zircaloy-4.4.