• Proceedings of the Korean Radioactive Waste Society Conference
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• 2012.05a
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• pp.195-196
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• 2012

• Nuclear Engineering and Technology
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• v.41 no.2
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• pp.163-170
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• 2009

In order to evaluate high-temperature oxidation behavior of the advanced alloy cladding under LOCA conditions, isothermal oxidation tests in steam were performed with cladding specimens prepared from high burnup PWR fuel rods that were irradiated up to 79 MWd/kg. Cladding materials were $M5^{(R)}$ and $ZIRLO^{TM}$, which are Nb-containing alloys. Ring-shaped specimens were isothermally oxidized in flowing steam at temperatures from 1173 to 1473 K for the duration between 120 and 4000s. Oxidation rates were evaluated from measured oxide layer thickness and weight gain. A protective effect of the preformed corrosion layer is seen for the shorter time range at the lower temperatures. The influence of pre-hydriding is not significant for the examined range. Alloy composition change generally has small influence on oxidation in the examined temperature range, though $M5^{(R)}$ shows an obviously smaller oxidation constant at 1273 K. Consequently, the oxidation rates of the high burnup $M5^{(R)}$ and $ZIRLO^{TM}$ cladding are comparable or lower than that of unirradiated Zircaloy-4 cladding.

• Proceedings of the KSME Conference
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• 2004.11a
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• pp.279-284
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• 2004

In this study, we suggested the ring creep test using the ring specimen of Arsene for estimating the burst creep properties of the cladding in stead of burst creep test. For this objective, we used the load-displacement conversion relationship of ring specimen called LCRR which had been determined on our previous study at high temperature by performing the ring tensile test and the numerical analysis. Then we carried out both the ring creep test and the burst creep test between 350 $^{\circ}C$ and 600$^{\circ}C$ which were higher then the in-service temperature of the cladding in a reactor. The creep properties from the ring creep test with applying LCRR were compared with those from the burst creep test of closed-end specimens. From the results, it could be seen an very strong relationship between them, especially in Larson- Miller parameter. So, it is expected that we can easily anticipate the creep properties of not only claddings but also various small pressure pipes using the ring creep test.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.29 no.7 s.238
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• pp.964-969
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• 2005

In this study, we suggested the ring creep test using the ring specimen of Arsene for estimating the burst creep properties of the cladding in stead of burst creep test. For this objective, we used the load-displacement conversion relationship of ring specimen called LCRR which had been determined on our previous study at high temperature by performing the ring tensile test and the numerical analysis. Then we carried out both the ring creep test and the burst creep test between $350^{\circ}C$ and $600^{\circ}C$ which were higher than the in-service temperature of the cladding in a reactor. The creep properties from the ring creep test with applying LCRR were compared with those from the burst creep test of closed-end specimens. From the results, it could be seen an very strong relationship between them, especially in Larson-Miller parameter. So, it is expected that we can easily predict the creep properties of not only claddings but also various small pressure pipes using the ring creep test.

• Korean Journal of Materials Research
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• v.29 no.8
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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.

• Proceedings of the Korean Nuclear Society Conference
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• 1995.05a
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• pp.611-616
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• 1995

핵연료 피복관의 일차 결함을 통해서 유입되는 냉각수에 의한 피복관 내면의 산화와 이에 따른 수소침투가 핵연료 피복관의 기계적 건전성에 미치는 영향을 규명하기 위한 연구를 수행하였다. 시험 시편은 Westinghouse, NRG, Sandvik에서 제조되는 Zircaloy-4 tube와 Westinghouse사에 개발한 신 합금인 ZIRLO™를 동일한 조건에서 수증기 산화와 수소 주입 실험을 수행하여 제조회사별 성능 평가를 하였으며 기계적 건전성 저하의 평가 방법으로 튜브 파열 실험(tube burst test)을 상온에서 수행하였다. 그 결과는 수소 침투량에 따라 피복관의 기계적 건전성이 지수적으로 감소하는 경향을 보였으며 500ppm이상에서는 취성파괴현상을 보이며 심각한 연성저하를 나타냈다. 제조회 사별 성능 평가에서는 A사 제품이 내식성과 수소흠수특성에서 다른 B, C, D사 제품에 비해 떨어지는 것으로 나타났다.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.05b
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• pp.183-188
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• 1997

핵연료 피복관용 Zr신합금을 개발하기 위해서 16종의 신합금을 설계하였다. 설계된 합금은 진공아크용해, $\beta$-열처리, 열간압연, 냉간압연 및 진공열처리의 공정에 의해 판재로 제조되었으며 이들 시편에 대해 35$0^{\circ}C$와 40$0^{\circ}C$에서 부식시험, 상온과 고온에서 인장시험 및 40$0^{\circ}C$에서 크립시험을 실시하여 신합금의 특성을 평가하였다. Zr-Nb-Sn계에 Fe, V, Te, Sb, Ru, Pd의 다른 원소를 미량 첨가하는 다원계 합금에서 Fe와 Cr은 부식특성을 향상시키는데 매우 효과적인 것으로 나타났다. Sb는 기계적강도를 향상시키고 Fe, Cr원소는 연신율을 증가시키는 원소로 밝혀졌으며 Sb와 V은 크립저항성을 매우 향상시킨다. 16종의 합금중 2-3종의 합금은 기존의 Zircaloy-4보다 우수한 내식성을 보였으며 Zr-Nb-Sn-FeCr합금은 ZIRLO와 유사한 부식저항성을 나타냈다. 부식과 크립저항성을 동시에 향상시키기 위해서는 Fe, Cr, Sb원소를 적절히 함유시킨 합금에 대해서 집중적인 연구가 수행되어야 할 것으로 사료된다.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.17 no.2
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• pp.127-137
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• 2019

Zr electrorefining is demonstrated herein using Zirlo tubes in a chloride-fluoride mixed molten salt in the presence of $AlF_3$. Cyclic voltammetry reveals a monotonic shift in the onset of metal reduction kinetics towards positive potential and an increase in intensity of the additional peaks associated with Zr-Al alloy formation with increasing $AlF_3$ concentration. Unlike the galvanostatic deposition mode, a radial plate-type Zr growth is evident at the top surface of the salt during Zr electrorefining at a constant potential of -1.2 V. The diameter of the plate-type Zr deposit gradually increases with increasing $AlF_3$ concentration. Scanning electron microscopy-energy-dispersive X-ray spectroscopy (SEM-EDX) and X-ray photoelectron spectroscopy (XPS) analyses for the plate-type Zr deposit show that trace amount of Al is incorporated as Zr-Al alloys with different chemical compositions between the top and bottom surface of the deposit. Addition of $AlF_3$ is effective in lowering the residual salt content in the deposit and in improving the current efficiency for Zr recovery.

• Nuclear Engineering and Technology
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• v.50 no.2
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• pp.229-236
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• 2018

Accident-tolerant fuels (ATFs) are currently of high interest to researchers in the nuclear industry and in governmental and international organizations. One widely studied accident-tolerant fuel concept is multilayer cladding (also known as coated cladding). This concept is based on a traditional Zr-based alloy (Zircaloy-4, M5, E110, ZIRLO etc.) serving as a substrate. Different protective materials are applied to the substrate surface by various techniques, thus enhancing the accident tolerance of the fuel. This study focuses on the results of testing of Zircaloy-4 coated with pure chromium metal using the cold spray (CS) technique. In comparison with other deposition methods, e.g., Physical vapor deposition (PVD), laser coating, or Chemical vapor deposition techniques (CVD), the CS technique is more cost efficient due to lower energy consumption and high deposition rates, making it more suitable for industry-scale production. The Cr-coated samples were tested at different conditions ($500^{\circ}C$ steam, $1200^{\circ}C$ steam, and Pressurized water reactor (PWR) pressurization test) and were precharacterized and postcharacterized by various techniques, such as scanning electron microscopy, Energy-dispersive X-ray spectroscopy (EDX), or nanoindentation; results are discussed. Results of the steady-state fuel performance simulations using the Bison code predicted the concept's feasibility. It is concluded that CS Cr coating has high potential benefits but requires further optimization and out-of-pile and in-pile testing.

• Korean Journal of Metals and Materials
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• v.49 no.1
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• pp.17-24
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• 2011

To investigate the relationship between heat treatment in zirconium alloy tubing process and metallurgical characteristics of Zr-1Nb-0.69Sn-0.11Fe alloy tubes, mechanical and oxidation behaviors of tubes heat treated at different temperatures after the final pilgering were investigated. The stress strain curves exhibited the saturation behaviors in all heat treatment conditions ($460{\sim}600^{\circ}C$) in this study with the onset strain of saturation increased with increase of post-pilgering annealing temperature. The strength fell off rapidly with increasing annealing temperature. The ultimate strength of the low tin Zr-1Nb-0.69Sn-0.11Fe alloy with slightly higher iron and oxygen contents in this study was found to be higher than Zr-1Nb-1Sn-0.1Fe alloy. The oxidation experiments in steam condition revealed that the corrosion resistance of low tin Zr-1Nb-0.69Sn-0.11Fe alloy was better than the Zr-1Nb-1Sn-0.1Fe alloy with a higher Sn content. The weight gain of low tin Zr-1Nb-0.69Sn-0.11Fe alloy tubes gradually increased with the increasing annealing temperature possibly due to the decreased Nb content in the matrix because of the formation of ${\beta}-Nb$ particles.