Korean Journal of Materials Research, Vol.11, No.8, 647-654, August, 2001
Zr-Nb계 합금의 석출물 특성과 산화 특성에 미치는 마지막 열처리 온도의 영향
Effect of Final Annealing Temperature on Precipitate and Oxidation of Zr- Nb Alloys
초록
Nb 첨가 Zr합금인 Zr-lNb합금과 Zr-lNb-lSn-0.3Fe함금의 석출물 및 산화 특성에 미치는 마지막 열처리 온도의 영향을 알아보기 위하여 최종 열처리 온도를 450 ? C 에서 800 ? C 까지 변화시켜 미세조직 및 산화 특성을 조사하였다. 부식 시험은 400 ? C , 수중기 분위기에서 270 일 동안 실시하였으며 X-선 회절법을 이용하여 산화막 결정 구조를 분석하였다. 마지막 열처리 온도가 600 ? C 이상일 때 두 합금 모두 β -Zr이 관찰되었으며 모두 재결정 이후 마지막 열처리 온도가 상승할수록 석출물의 면적 분율이 증가하는 경향을 나타내었다. 모든 열처리 온도 구간에서 Zr-lNb합금의 부식 저항성이 Zr-lNb-lSn-0.3Fe 합금에 비해 우수하였으며 두 합금 모두 재결정 이후 부식 저항성이 급격히 나빠졌다. 이는 600 ? C 이후 형성된 β -Zr의 영향으로 밝혀졌다.
ffects of final annealing temperature on the precipitate and oxidation were investigated for the Zr-lNb and Zr-lNb-lSn-0.3Fe alloys. The microstructure and oxidation of both alloys were evaluated for the optimization of final annealing process of these alloys in the annealing temperature regime of 450 to 800 ? C . The corrosion test was performed under steam at 400 ? C for 270 days in a static autoclave. The oxide formed was identified by low angle X-ray diffraction method. The β -Zr was observed at annealing temperature above 600 ? C . Above 600 ? C , the precipitate area volume fraction of Zr-lNb and Zr-1Nb-lSn-0.3Fe alloys appeared to be increased with increasing the final annealing temperature. The corrosion resistance of Zr-lNb was higher than that of Zr- lNb-lSn-0.3Fe alloy. The corrosion rate of both alloys were accelerated due to the formation and growth of β -Zr with increasing the annealing temperature.
- Baek JH : KAERI Report, KAERI/AR-547/99, 17 (1999) (1999)
- Garzaroli F : Zirconium in the Nuclear Industry, ASTM STP 1132, 35 (1991) (1991)
- Isobe T, Matsuo Y, Zirconium in the Nuclear Industry, ASTM STP 1132. 125 (1991) (1991)
- Sabol GP, Kilp GR, Balfour MG, Roberts E, Zirconium in the Nuclear Industry. ASTM STP 1023. 227 (1989) (1989)
- Comstock RJ, Schoenberger G, Sabol GP, Zirconium in the Nuclear Industry. ASTM STP 1295, 710 (1996) (1996)
- Marden JP, Charquet D, Senevat J, Zirconium in the Nuclear Industry. ASTM STP 1354. 357 (1998) (1998)
- Wadekar SL, Banerjee S, Raman VV, Asundi MK, Zirconium in the Nuclear Industry, ASTM STP 1132, 140 (1991) (1991)
- Knorr DB, Notis MR, J. Nucl. Mater., 18 (1975)
- Nomura S, Akutsu C, Electrochem. Techno., 4, 198 (1989)
- Schemel JH : Zirconium alloy fuel clad tubing engineering guide, Sandvik Special Metals, Kennewick, WA., 298 (1989) (1989)
- ASTM-G2 : Standard Test Method for Corrosion Testing of Products of Zirconium, Hafnium and Their Alloys in Water at 680 ? F or in Steam at 750 ? F
- Kim KH, Choi BK, Baek JH, Kim SJ, Jeong YH, J. Kor. Inst. Met & Mater., 9, 188 (1999)
- Holm K, Embury JD, Acta Metallur., 25, 1191 (1977)
- Lee MH, Kim HG, Yoon YG, Jeong YH, J. Kor. Inst. Met. & Mater., 7, 923 (2000)
- Kass S, Corrosion-NACE, 27, 443 (1971)
- Jeong YH, Korean J. Mater. Res., 6, 585 (1996)
- Kim HG, Lim YS, Wey MY, Jeong YH, J. Kor. Inst. Met & Mater., 37, 584 (1999)
- Zaimovsky AS, Nikulina AV, Reshetnikov NG, Zr Alloys In Nuclear Power, Moscow, Energoizdat, (1981) (1981)
- Godlewsky J, Zirconium in the Nuclear Industry, ASTM STP 1245, 663 (1994) (1994)
- Godlewsky J, Zirconium in the Nuclear Industry, ASTM STP 1132, 416 (1991) (1991)