Rupture of a cold-worked Zircaloy-2 pressure tube in Pickering Unit 2 in 1983 occurred when a crack developed from an array of hydride blisters. These had formed on the outside surface of the pressure tube where it contacted the surrounding calandria tube. Fol-lowing this event, a comprehensive research program was initiated by Ontario Hydro. Its objective was to determine the consequences of pressure tube-calandria tube contact for channels in other reactors with Zr-2.5 Nb pressure tubes. The purpose of this paper is to present some of the results that have been obtained in the areas of blister growth, fracture, and nondestructive evaluation and to show how these results may be applied to Canada Deuterium Uranium (CANDU) reactor inspection. Blisters were grown both on small specimens and on pressurized tubes. Blisters were ob-served to grow by gradual precipitation of hydride platelets throughout a volume that also increased with time. Blister growth rates increased rapidly with increasing average temperature. The hydride distribution in the blisters has been observed to be highly variable. Four-point bending fracture tests were carried out on the small specimen blisters. These fracture tests indicate that blister size is a key parameter in determining the stress required to crack the blister with larger blisters cracking at lower applied stresses. This trend is shown to be consistent with previous finite-element modelling, which incorporated hydride plasticity. Two of the pressurized tubes were inspected with a specially designed ultrasonic head, and responses were correlated with metallographic information. The ultrasonic inspection system was capable of detecting cracks as small as 0.15 mm deep in small blisters. This gives some expectation that inspection will be able to identify blistered tubes.
Description This unique publication provides an international overview on the production and use of Zr alloys, their properties and behavior during nuclear service, the design of Zr components and their testing after service. Papers cover the historical aspects of research on Zr alloys; basic metallurgy, including studies of second phase particles; irradiation creep and growth; material performance during LOCA (loss of coolant accidents); and RIA (reactivity initiated accident). Half of the papers in STP 1423 relate to corrosion and hydriding behavior, the most important current issues in the industry. A quarter of the papers deal with in-reactor studies. The remaining papers discussed the behavior and properties of Zr alloys for the intermediate storage of spent fuel. STP 1423 is a valuable resource for engineers and scientists involved in the processing and properties of Zr alloys, the production and use of Zr alloy reactor components, their behavior during service and property changes that occur with increasing neutron fluences. 42 peer-reviewed papers provide an international overview of the production and use of Zr alloys. Learn about: • Production and use of Zr alloys in the nuclear industry • Properties and behavior during nuclear service • Corrosion and hydriding during service • Dimensional changes during service • Design of Zr components • Testing after service • Effect of alloy microstructure and composition on behavior • Studies of the composition, properties and behavior of the oxides formed during service
The quality of the papers in this publication reflects not only the obvious efforts of the authors and the technical editor(s), but also the work of the peer reviewers. In keeping with long standing publication practice, ASTM maintains the anonymity of the peer reviewers. The ASTM Committee on Publications acknowledges with appreciation their dedication and contribution of time and effort on behalf of ASTM.
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