Zirconium in the nuclear industry: Eleventh international symposium

Conference · Tue Dec 31 04:00:00 EST 1996

OSTI ID:479426

Bradley, E R ^[1]; Sabol, G P ^[2]

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1. ed.; Sandvik Special Metals Corp., Kennewick, WA (United States)
2. ed.; Westinghouse Electric Corp., Pittsburgh, PA (United States)

Since their development in the 1950s and introduction into commercial nuclear power plants in the 1960s, the zirconium-based alloys Zircaloy-2 and -4, Zr-1Nb, and Zr-2.5Nb are the alloys currently used in the world`s reactors. However, with increasing fuel duty, the margins displayed by these alloys have eroded, and considerable research has been conducted to improve these materials and also to develop more advanced alloys. Optimization of alloying constituents and processing parameters, coupled with a more basic understanding of performance-limiting phenomena, are the primary themes of most of the papers contained herein. Fully half of the papers are directly concerned with the corrosion of Zr-based alloys. The detailed characterization of the effects of irradiation on the microstructure of irradiated Zircaloys has confirmed the loss of iron from second phase particles, with or without amorphization of the particles. Also, the correspondence between iron in solution in the matrix, the formation of -type dislocations, and the onset of accelerated irradiation induced growth has been verified. Unfortunately, the role of dissolved iron in the matrix in the nucleation of -type dislocations has not been established. One observation that has been verified, however, is the low irradiation growth in Zr-Nb-Sn-Fe alloys, also presumably due to suppression of dislocation formation. Several papers in the symposium focused on fuel clad modeling, and although a schism still exists between fundamental material properties and fuel performance predictive codes, the modeling papers presented are attempts to link component response to the quantifiable material behavior in a manner consistent with qualitative structural observations. In summary, the data, analyses, hypotheses, and theories presented in this book represent the current state of zirconium technology as applied to nuclear power reactors. Separate abstracts were prepared for 43 papers in this volume.

• OSTI ID: 479426
• Report Number(s): CONF-950926--
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
36 MATERIALS SCIENCE
CORROSION
HYDROGEN EMBRITTLEMENT
LEADING ABSTRACT
MECHANICAL PROPERTIES
MEETINGS
MICROSTRUCTURE
NUCLEAR POWER PLANTS
PHYSICAL RADIATION EFFECTS
REACTOR MATERIALS
ZIRCONIUM ALLOYS