• Proceedings of the Korean Nuclear Society Conference
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• 1996.05c
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• pp.341-346
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• 1996

사용후핵연료에 대한 장기건식저장과 관련하여 원자로에서 조사된 사용후 핵연료피복관에 대한 산화시험을 공기분위기에서 수행하였다. 피복관 시료의 50$0^{\circ}C$ 공기중 산화시험 결과 산화 초기에 급격한 산화율을 보였으며, 이 후 천이점까지 느리게 산화가 진행되다가 천이 후에는 선형적으로 급격히 무게가 증가하는 지르코늄 합금의 수증기 및 공기중에서의 전형적인 산화양상을 나타내었다. 시편별로는 가장 두꺼운 노내 산화막을 가진 시편이 가장 높은 산화율을 나타내었으며, 노내 산화시 천이점에 근접한 시편들이 가장 낮은 산화율을 보였다. 산화율이 가장 높은 시편의 천이후 영역에서의 산화율은 $\Delta$W ＝ 0.74 t ＋ 38.61과 같은 관계식으로 표현될 수 있었다. 이 때 $\Delta$W는 무게이득(mg/dm$^2$)이고 t는 산화시간(h)을 나타낸다. 시험에 사용된 피복관의 단위 산화막두께(l$\mu$m)에 대한 산화무게증가량은 약 13.4mg/dm$^2$으로 나타났다. 이러한 결과들은 사용후핵연료 중간저장 시설 및 저장캐스크의 설계 전산코드 작성 및 저장시설의 운영에 관련되어 기반자료로 활용될 수 있을 것이다.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05c
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• pp.238-243
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• 1996

현재 상용 공급되고 있는 4종의 경수로 핵연료 피복관에 대해 노외 부식특성 시험을 수행하여 그 특성을 비교하였다. 이중 3종의 피복관은 제조공정을 달리하여 제조된 low tin Zircaloy-4 (피복관 A, B, C)이며, 1종은 Nb이 첨가된 Zr 합금 (피복관 D)이었다. 증류수내 Li 함량을 2.2 ppm, 30 ppm, 220 ppm으로 변화시키며 부식시험을 수행한 결과 최종 pilgering시 낮은 Q 인자와 높은 열처리 온도로 제조된 피복관 A의 내부식성이 대체로 우수하였으며, 220 ppm Li 수용액에서는 Nb이 첨가된 피복관의 내부식성이 매우 우수한 것으로 나타났다. Li 첨가의 영향을 보면 2.2 ppm Li 첨가시에는 증류수와 거의 동일한 부식거동을 나타내고 있으나 30 ppm Li 첨가시에는 부식이 가속되고 있었으며, 220 ppm Li으로 Li 함량이 크게 증가하였을 때 부식속도도 크게 증가하였다. 수소흡수율은 피복관 A에서 가장 높았으며, 피복관 D가 가장 낮은 값을 나타내었다.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05c
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• pp.259-264
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• 1996

상용 Zircaloy-4보다 성능이 우수한 고연소도용 Zr 신합금을 개발하는 것을 목표로 외국에서 개발중인 12종의 신합금 피복관에 대한 특성평가, 부식기구 규명 연구, 국내에서 제조된 Zr 신합금의 특성평가를 실시하였다. 외국 피복관의 부식특성 평가로 부터 Sn을 0.6-1.0 wt.% 첨가하고 Nb을 0.4 wt.% 첨가하는 것이 내식성 관점에서 바람직함을 알 수 있었다. 여러 가지 LiOH용액에서의 부식기구 연구를 통해 수소화물이 부식가속의 원인임을 알 수 있었으며 수소화물 형성을 억제하는데는 Nb첨가가 효과적인 것으로 나타났다. 이와 같은 연구결과를 토대로 신합금의 개발방안을 수립하였으며 예비적으로 합금을 설계. 제조하여 특성시험을 실시한 결과, Zr-Sn-Nb-FeCr 합금이 우수한 내식성을 보이며 Fe, Mo는 강도 증가 효과가 큰 것으로 나타났다. 이러한 연구결과를 종합적으로 평가하여 신합금을 설계하고 노외성능 평가를 통해서 신합금을 선정한후, 단계적으로 하나로를 이용한 노내성능 평가를 실시할 예정이다.

• Journal of the Korean Society for Heat Treatment
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• v.17 no.4
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• pp.216-222
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• 2004

Study was focused on the quenching embrittlement property of Zircaloy-4 cladding simulated Loss Of Coolant Accident (LOCA) environment in terms of high temperature oxidation and phase transformation. Property in LOCA condition of advanced cladding that contained Nb element was also investigated. Claddings were oxidized at given temperature and given time followed by water quenching. The results showed that ${\beta}$ phase which formed at quenching stage has an influence on cladding property. In case of advanced cladding, Nb retards cladding oxidation, thus enhances quenching resistance.

• Nuclear Engineering and Technology
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• v.24 no.1
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• pp.75-85
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• 1992

The texture changes and their effect on the 0.2% yield strength of Zircaloy-4 sheet were examined after quenched from the ($\alpha$+$\beta$) phase temperature. When the prior ($\alpha$+$\beta$) gram size was slightly larger than that of the $\alpha$-annealed, the observed texture was similar to the $\alpha$-annealed texture having an ideal orientation of the (0001) basal pole at 30$^{\circ}$away from the normal direction toward the transverse direction. When the prior ($\alpha$+$\beta$) grain size was twice as large as that of the $\alpha$-annealed, the location of maximum basal pole intensity was distributed between the transverse and the rolling direction making an angle 15$^{\circ}$from the normal direction, and the observed texture became isotropic. It was found that the Kearns texture parameter, fr, in the rolling direction increased steadily, and fr in the transverse direction increased slightly, while fr in the the normal direction decreased with increasing heat treatment time. With a small increase in fr, the 0.2% yield strength increased drastically. The influence of texture was analyzed by deriving the Schmid orientation factors and the resolved shear stresses for the deformation systems. It was found that the large increase in the 0.2% yield strength was attributed mainly to the microstructural changes and partly to the texture changes by the ($\alpha$+$\beta$) heat treatment.

• Nuclear Engineering and Technology
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• v.18 no.4
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• pp.267-272
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• 1986

To obtain various necessary data for the manufacturing and the use of the nuclear fuel cladding tube, the effects of deformation and heat treatment on Properties of Zircalof-4 material have been studied. The hardness is increased rapidly at a low degree of cold work and increased rapidly at cold work above 10%. Recrystallization has been completed at 64$0^{\circ}C$, 59$0^{\circ}C$, and 555$^{\circ}C$ in 30%, 60% and 80% cold worked specimen, respectively. The transformation of microstructure with increasing cooling rate after $\beta$-annealing is as follows; coarse Widmanstatten ($\alpha$) longrightarrow fine parallel plate ($\alpha$) longrightarrow martensite ($\alpha$$^{'}$). At the same time, hardness increased with increasing cooling rate. rate.

• Korean Journal of Materials Research
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• v.14 no.8
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• pp.579-586
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• 2004

A study on the effects of oxide growth on the mechanical properties degradation of pure zirconium and Zircaloy-4 is carried out with high temperature tensile tests. It is found that the mechanical properties can deteriorate with the oxide growth less than $1\%$ of total specimen cross section, especially at $300\~400^{\circ}C$ that is zirconium alloy cladding temperature during the nuclear reactor operation. It is also revealed that Young's modulus changes little but yield strength and tensile strength drop down to $20\% and 40\%$ of the room temperature strength, respectively, in the temperature range. Fractographic analysis shows that the number of dimples decreases and fractured surface becomes smooth with increasing oxide thickness.

• Journal of Welding and Joining
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• v.21 no.6
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• pp.20-25
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• 2003

Various welding methods such as Gas Tungsten Arc Welding(GTAW), magnetic force electrical resistance welding and Laser Beam Welding(LBW) are now available for end cap closure of nuclear fuel rods. Even though the resistance and GTA welding processes are widely used in manufacturing commercial fuel rods, they can not be recommended for the remote seal welding of fuel rods in the hot cell Facility due to the complexity of the electrode alignment, the difficulty in replacing parts in a remote manner and the large heat input for the thin sheath. Therefore, the Nd:YAG laser system using optical fiber transmission was selected for the end cap welding of irradiation fuel rods in the hot cell. The remote laser welding apparatus in the hot cell Facility was developed using a pulsed Nd:YAG laser of 500 watt average power with an optical fiber transmission. The weldment quality such as microstructure and mechanical strength was satisfactory. The optimum conditions of laser welding for encapsulating irradiation fuel rods in the hot cell were obtained.

• Journal of Welding and Joining
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• v.19 no.1
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• pp.88-94
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• 2001

DUPIC (Direct use of spent PWR fuel in CANDU reactors) nuclear fuel is a CANDU fuel fabricated remotely from spent PWR fuel materials in a hot cell. The soundness of the end closure welds of nuclear fuel elements is an important factor for the safety and performance of nuclear fuel. To evaluate the soundness of the end closure welds of DUPIC fuel element, a precise X-ray inspection system is developed using a micro-focus X-ray generator with an image intensifier and a real time camera system. The fuel elements made of Zircaloy-4 and stainless steel by an Nd:YAG laser welding and a TIG welding aye inspected by the developed inspection system. The soundness of the welds of the fuel elements was confirmed by the X-ray inspection process, and the irradiation test of DUPIC fuel elements has been successfully completed at the HANARO research reactor.

• Proceedings of the Korean Nuclear Society Conference
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• 1999.05a
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• pp.279-279
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• 1999

To improve the corrosion model of the fuel performance analysis code COSMOS, a model was developed considering thermohydraulic phenomena and the effect of water chemistry and low Sn in the alloy composition on the corrosion behavior. It is assumed that the lithium enhancement factor influences the corrosion behavior only if the subcooled void is present in the coolant. The developed model was verified with the database obtained from Grohnde and Ringhals 3 reactors. Comparison of predicted oxide thickness with measured data showed the applicability of COSMOS code to analyze the cladding oxidation. In the future, the effect of the hydride in the cladding and the precipitate changes due to irradiation should be included.cluded.