• 한국재료학회지
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• 제14권8호
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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.

• 한국진공학회지
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• 제4권S2호
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• pp.100-105
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• 1995

Nitrogen implantation process has been applied for improvement of wear resistance of Zircaloy-4 fuel cladding materials. Nitrogen was implanted at 120keV to a total dose range of $1\times 10^{17}$ions/$\textrm{cm}^2$ to $1\times 10^{18}$ions/$\textrm{cm}^2$ at various temperatures between $270^{\circ}C$ and $671^{\circ}C$. The microstructure changes by nitrogen implantation were analyzed by XRD and AES and wear behavior was evaluated by performing ball-on-disc type wear testing at various loads and sliding velocities under unlubricated condition. Nitrogen implantation produced ZrNx nitride above $3\times 10^{17}$ions/$\textrm{cm}^2$ as well as heavy dislocations, which resluted in an increase in microhardness of the implanted surface of up to 1400 $H_k$ from 200 $H_k$ of unimplanted substrate. Hardness was also found to be increased with increasing implantation temperature up to 1760 $H_k$ at $620^{\circ}C$. The wear resistance was greatly improved as total ion dose and implantation temperature increased. The effective enhancement of wear resistance at high dose and temperature is believed to be due to the significant hardening associated with high degree of precipitation of Zr nitrides and generation of prismatic dislocation loops.

• Nuclear Engineering and Technology
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• 제21권2호
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• pp.89-98
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• 1989

지르칼로이 -4 피복관에 대해서 3가지 변형 속도로(1.2$\times$10E-7/s, 2.0$\times$10E-6/s, 3.2$\times$10E -5/s), 553-873K의 온도 구간에서 구리 맨드렐 팽창 시험법을 공기와 진공(5$\times$10E-5 torr) 분위기에서 수행했고, 변형 속도의 변화는 시편의 가열 속도를 조절함으로써 얻을 수 있었다. 각각의 변형 속도에서 항복 응력 피크와 변형 속도 감도 최저값 그리고 활성화 부피 극대값이 나타나는 이유는 동적변형시효 현상 때문이라고 설명된다. 항복 응력 피크가 나타나는 온도와 변형속도로부터 얻어진 동적변형시효의 활성화 에너지는 196(KJ/mol)이었고 이 값은 $\alpha$-지르코니움과 지르칼로이-2에서 활성화 에너지(207-220 KJ/mol)값과 잘 일치한다. 그러므로 573-673K의 온도 구간에서 나타나는 동적변형시효 현상은 산소원자 때문이라고 생각된다. 산화에 의한 항복 응력의 증가는 공기중 실험과 진공 실험으로 얻어진 항복 응력값을 비교함으로서 얻었고, 그것은 항복 응력의 증가 분율로 표시했다. 결과는 변형속도가 느릴 수록 증가 분율은 더욱 더 커짐을 알 수 있었다. 그리고 산소 침투량과 항복 응력 증가 사이의 관계가 직선적이라는 가정하에 공기와 수중에서의 산화 속도를 비교하여 수중에서의 항복 응력 값을 계산해 보았다.

• 한국재료학회지
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• 제18권6호
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• pp.326-333
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• 2008

Since the 1990s, the second generation of Zirconium alloys containing main alloy compositions of Nb, Sn and Fe have been used as a replacement of Zircaloy-4 (Zr-Sn-Fe-Cr), a first-generation Zirconium alloy, to meet severe and rigorous reactor operating conditions characterized by high-burn-up, high-power and high-pH operations. In this study, the mechanical properties and creep behaviors of Zr-Sn-Fe-Cr and Zr-Nb-Sn-Fe alloys were investigated in a temperature range of $450{\sim}500^{\circ}C$ and in a stress range of $80{\sim}150\;MPa$. The mechanical testing results indicate that the yield and tensile strengths of the Zr-Nb-Sn-Fe alloy are slightly higher compared to those of Zr-Sn-Fe-Cr. This can be explained by the second phase strengthening of the $\beta$-Nb precipitates. The creep test results indicate that the stress exponent for the steady-state creep rate decreases with the increase in the applied stress. However, the stress exponent of the Zr-Sn-Fe-Cr alloy is lower than that of the Zr-Nb-Sn-Fe alloy in a relatively high stress range, whereas the creep activation energy of the former is slightly higher than that of the latter. This can be explained by the dynamic deformation aging effect caused by the interaction of dislocations with Sn substitutional atoms. A higher Sn content leads to a lower stress exponent value and higher creep activation energy.

• 한국윤활학회:학술대회논문집
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• 한국윤활학회 2003년도 학술대회지
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• pp.64-71
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• 2003

In order to evaluate the effect of alloying elements (Nb and Sn) on the wear resistance of advanced Zr fuel claddings, sliding wear tests have been performed in room temperature air and water and these results were compared with those of commercial alloys such as Zircaloy-4, A and B alloys. As a result, the advanced Zr fuel claddings have a similar wear resistance compared with the commercial alloys. The wear resistance of the advanced Zr fuel claddings is closely releted to the content of Nb and Sn even though the effects of transition elements are involved in deforming wear properties. In the tested specimens with similar Sn content, wear volume became down to a minimum at $0.4\;wt\;\%$ Nb, then rapidly increased at 1.0 wt Nb. This behavior results in the variation of grain size with alloying contents. But Sn did not have a significant effect on the wear volume of advanced Zr fuel claddings below $1.1\;wt\%$. The relationship between alloying elements and wear behaviour was evaluated and discussed using material compatibility factor.

• Journal of Welding and Joining
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• 제7권2호
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• pp.60-69
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• 1989

The present study was aimed at investigating the effect of welding parameters such as welding current, electrode force(or squeeze force) and parts cleaning on the sound weld, and establishing the most reliable weld conditions for HWP(Heavy Water Reactor) fuel end capping with the resistance upset butt welding. Major results obtained are as follows. 1. The amount of sound weld was increased with increasing weld current(5.0-11KA) because the activated diffusion with increasing heat generation played an important role in eliminating the porosity and weld line in the weld interface. 2. It was found that weld current was not significantly influenced by the electrode force although the increase of it caused a slight increase of weld current and upset deformation. 3. Acetone rinsing before drying for the Zircaloy-4 end cap cleaning produced the reliable sound weld because it would remove the remaining solvent and surface films, and provided the uniform contact between the end cap and the tube. 4. The optimum welding conditions for fuel end capping by a resistance upset hytt welding are obtained as follows. weld current: 10-11KA, electrode force: 62-90KPa parts cleaning: vapor degreasing.rarw.water, acetone rinsing.rarw.drying.

• Nuclear Engineering and Technology
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• 제40권1호
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• pp.21-36
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• 2008

High bum-up fuel technology has been developed through a national R&D program, which covers key technology areas such as claddings, $UO_2$ pellets, spacer grids, performance code, and fuel assembly tests. New cladding alloys were developed through alloy designs, tube fabrication, out-of-pile test and in-reactor test. The new Zr-Nb tubes are found to be much better in their corrosion resistance and creep strength than the Zircaloy-4 tube, owing to an optimized composition and heat treatment of the new Zr-Nb alloys. A new fabrication technology for large grain $UO_2$ pellets was developed using various uranium oxide seeds and a micro-doping of Al. The uranium oxide seeds, which were added to $UO_2$ powder, were prepared by oxidizing and heat-treating scrap $UO_2$ pellets. A $UO_2$ pellet containing tungsten channels was fabricated for a thermal conductivity enhancement. For the fuel performance analysis, new high burnup models were developed and implemented in a code. This code was verified by an international database and our own database. The developed spacer grid has two features of contoured contact spring and hybrid mixing vanes. Mechanical and hydraulic tests showed that the spacer grid is superior in its rodsupporting, wear resistance and CHF performance. Finally, fuel assembly test technology was also developed. Facilities for mechanical and thermal hydraulic tests were constructed and are now in operation. Several achievements are to be utilized soon by the Korea Nuclear Fuel and thereby contribute to the economy and safety of PWR fuel in Korea

• 한국재료학회:학술대회논문집
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• 한국재료학회 2003년도 추계학술발표강연 및 논문개요집
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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.

• 한국세라믹학회지
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• 제48권1호
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• pp.69-73
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• 2011

A $ZrO_2$ thin film as a corrosion protective layer was deposited on Zircaloy-4 (Z-4) clad material using $N_2O$ as a reactive gas by RF reactive magnetron sputtering at room temperature. The Z-4 substrate was located in plasma or out of plasma during the $ZrO_2$ deposition process to investigate mechanical and corrosive properties for the plasma immersion. Tetragonal and monoclinic phases were existed in $ZrO_2$ thin film immersed in plasma. We observed that a grain size of the $ZrO_2$ thin film immersed in plasma state is larger than that of the $ZrO_2$ thin film out of plasma state. In addition, the corrosive property of the $ZrO_2$ thin films in the plasma was characterized using the weight gains of Z-4 after the corrosion test. Compared with the $ZrO_2$ thin film immersed out of plasma, the weight gains of $ZrO_2$ thin film immersed in plasma were larger. These results indicate that the $ZrO_2$ film with the tetragonal phase in the $ZrO_2$ can protect the Z-4 from corrosive phenomena.

• Nuclear Engineering and Technology
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• 제55권7호
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• pp.2578-2590
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• 2023

In this work, a 24-month two-batch fuel management strategy for the APR1400 using LEU + has been investigated, where enrichments of 5.9 and 5.2 w/o are utilized in lieu of the conventional 4-5 w/o UO₂ fuel. In addition, an Accident Tolerant Fuel (ATF) clad based on the swaging technology is applied to APR1400 fuel assemblies. In this special ATF clad design, both outer and inner SS316 layers protect the conventional zircaloy clad. Erbia (Er₂O₃) is introduced as a burnable absorber with two-fold goals to lower the critical boron concentration in the long-cycle LEU + loaded core as well as to handle the LEU + fuel in the existing front-end fuel facilities without renewing the license. Two types of fuel assemblies with different loading of gadolinia (Gd₂O₃) are considered to control both the reactivity and the core radial power distribution. The erbia burnable absorber is uniformly admixed with UO₂ in all fuel pins except for the gadolinia-bearing ones. In this study, two core designs were devised with different erbia loading, and core performance and safety parameters were evaluated for each case in comparison with a core design without any burnable absorbers. The core analysis was done using the two-step method. First, cross-sections are generated by the SERPENT 2 Monte Carlo code, and the 3-D neutronic analysis is performed with an in-house multi-physics nodal code KANT.