The fuel cladding for pressurized and boiling water reactors is usually one of the zirconium-tin alloys, Zircaloy-2 or Zircaloy-4. These alloys have sufficient strength and corrosion resistance to withstand the rigors of reactor service as well as low neutron absorption characteristics which increases the neutron efficiency and permits a reduction in the fuel enrichment of a reactor. Operational experience has shown that Zircaloy fuel cladding performs satisfactorily under present service conditions.1  However, failures of experimental Zircaloy-clad fuel elements indicate that operational limitations need further definition before a Zircaloy-clad core is committed to the demanding conditions of some of the more advanced reactor designs.

Subject
Water, Fluxes, Corrosion rate, Heaters, Reactors, Autoclaves, Aqueous environments, Dissolved hydrogen, Hydrogen absorption, Flanges, Heat, Corrosion data, Hydrogen
© 1969 Association for Materials Protection and Performance (AMPP). All rights reserved. No part of this publication may be reproduced, stored in a retrieval system, or transmitted, in any form or by any means (electronic, mechanical, photocopying, recording, or otherwise) without the prior written permission of AMPP. Positions and opinions advanced in this work are those of the author(s) and not necessarily those of AMPP. Responsibility for the content of the work lies solely with the author(s).
1969
Association for Materials Protection and Performance (AMPP)
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