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The waterside nodular corrosion of Zircaloy cladding in boiling water reactors (BWRs) remains a primary concern in fuel performance because of (1) recurring fuel failures in several BWRs and (2) the current trend in the nuclear industry towards higher fuel burnups. The failures might be avoided if there were a reliable way of sorting claddings, so that only cladding that is highly resistant to corrosion would be used. In-reactor corrosion testing of cladding tubes is expensive and time-consuming, and consequently, there exists a strong incentive to develop a short-term, out-of-pile corrosion test that is able to predict cladding corrosion behavior for BWR applications.
It is shown in this paper that a high-pressure steam autoclave test performed at 520°C for 24 h is capable of classifying the in-reactor nodular corrosion properties of the fuel cladding. The effects of test temperature, sample surface treatment, and oxygen and hydrogen content in the steam on corrosion performance are documented.
Zircaloy tubing, corrosion tests, out-of-pile corrosion behavior, in-reactor behavior
Staff engineer, ABB ATOM AB, Västerås,
Program manager, Electric Power Research Institute, Palo Alto, CA