산화막 성장이 지르코늄 합금의 기계적 물성 열화에 미치는 영향

전상환; 김용수; Sang-hwan Jeon; Yong-soo Kim

Korean Journal of Materials Research, Vol.14, No.8, 579-586, August, 2004

DOI10.3740/MRSK.2004.14.8.579 Export Citation

산화막 성장이 지르코늄 합금의 기계적 물성 열화에 미치는 영향

Effects of Oxide Growth on Mechanical Properties Degradation of Zirconium Alloys

전상환, 김용수^†

한양대학교 원자력공학과

Sang-hwan Jeon, and Yong-soo Kim^†

Department of Nuclear Engineering, Hanyang University, Korea

E-mail:

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 ? 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%and40% 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.

Keywords:zirconium alloys;oxidation;mechanical properties degradation;yield strength;tensile strength

[References]

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[2] Yang R, Ozer O, Rosenbaum HS, Proc. Intern. Topical Meeting on LWR Fuel Performance (Park city, Utah, USA, April 2000). ed. Rosa Yang (American Nuclear Society, Illinois, USA, 1997) (1997)

[3] Pati SR, Jourdain P, Smith GP, Garde AM, Proc. Intern. Topical Meeting on LWR Fuel Performance (Portland, Oregon, USA, March 1997). ed. Painter CL (American Nuclear Society, Illinois, USA, 1997) p. 413 (1997)

[4] Etoh Y, Shimada S, Adamson RB, Yasuda T, Kogai T, Ishii Y, Proc. Intern. Topical Meeting on LWR Fuel Performance (Portland, Oregon, USA, March 1997). ed. Painter CL (American Nuclear Society, Illinois, USA, 1997) p. 211 (1997)

[5] USNRC Standard Review Plan, NUREG-75/087, Washington (1996) (1996)

[6] Cheng BC, Brown JM, Turnage KG, Armstrong EA, Hudson M, Proc. Intern. Topical Meeting on LWR Fuel Performance (Portland, Oregon, USA, March 1997). ed. Chad L. Painter (American Nuclear Society, Illinois, USA, 1997) p. 379 (1997)

[7] Johnson AB, Jr., Zirconium Alloy Oxidation and Hydriding Under Irradiation: Review of Pacific Northwest Laboratories' Test Program Results, EPRI NP-5132, Battelle, Pacific Northwest Laboratories, USA (1987) (1987)

[8] IAEA Co-ordinated Research Project on Establishment of a Thermo-physical Properties Data Base of Light Water Reactors and Heavy Water Reactors, IAEA-RC-734.3, Austria (2003) (2003)

[9] Kanashov B, Fuel Failure in Water Reactors: Causes and Mitigation, IAEA-TECDOC-1345, Slovakia (2002) (2002)

[10] Parsons PD, Hindle ED, Mann CA, PWR Fuel Behaviour in Design Basis Accident Conditions, CSNI Report 129, OECD Nuclear Energy Agency (1986) (1986)

[11] Sawatzky A, Zirconium in Nuclear Industry, ASTM STP, 681, 479 (1979) (1979)

[12] MATPRO - A Labrary of Materials Properties for Light-Water-Reactor Accident Analysis, NUREG/CR-6150, Idaho National Engineering Laboratory, USA (1993) (1993)

[13] Dressler G, Drefahl K, Matuda H, Wincierz P, Zirconium in Nuclear Applictions, ASTM STP, 551, 92 (1974) (1974)

[14] Baty D, Pavinch WA, Dietrich MR, Clevinger GS, Papazoglou TP, Zirconium in the Nuclear Industry ASTM STP, 824, 306 (1982)

[15] Brunstetter DR, Kling HP, Alexander BH, The Tensile Properties of Zirconium at Elevated Temperatures, p. 5, NYO-1126, United States Atomic Energy Commission, USA (1950) (1950)

[16] Simcoe CR, Mudge WL, Jr. Tensile Properties of Zirconium and Zirconium Alloys, p. 8, AECD-3680, United States Atomic Energy Commission, USA (1951) (1951)