A review of cladding failure thresholds in RIA conditions based on transient reactor test data and the need for continued testing

Conference · Fri Oct 14 04:00:00 EDT 2022

DOI:https://doi.org/10.13182/TopFuel22-38944· OSTI ID:2475185

Kamerman, David W ^[1]; Jensen, Colby B ^[1]; Folsom, Charles P ^[1]; Woolstenhulme, Nicolas E ^[1]; Wachs, Daniel M ^[1]

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1. Idaho National Laboratory

Transient reactor experiments on light water reactor (LWR) fuel pins had been conducted since the beginning of the nuclear era to help determine core coolability and cladding failure thresholds. During one such test in November of 1993 at the CABRI transient test reactor on a test involving a high burnup fuel rod with a corroded Zircaloy-4 cladding it was first observed that cladding failures could occur prior to a departure from nucleate boiling (pre-DNB) at lower-than-expected peak radial average enthalpies. This paper will present an independent review of the publicly available transient reactor test database on higher burnup LWR pins conducted at the CABRI and NSRR reactors as well as review of a selection of published out of pile mechanical testing methods. The purpose of the review is to determine how well the new regulatory limits are supported by experimental data. The review will identify if additional transient reactor tests could provide additional support for the NRC guidance or identify the need for revisions. The evaluation will consider how far the existing database can be extrapolated when considering low hydrogen zirconium alloy claddings (with and without protective coatings) containing very high burnup (> 70 MWd/kgU) UO2 fuel pellets. Finally, the authors will suggest how out of pile mechanical tests can be used in conjunction with a limited number of transient reactor tests to develop cladding specific failure thresholds in RIA type transients.

• Research Organization: Idaho National Laboratory (INL), Idaho Falls, ID (United States)
• Sponsoring Organization: 60
• DOE Contract Number: AC07-05ID14517
• OSTI ID: 2475185
• Report Number(s): INL/CON-22-67260-Rev000
• Country of Publication: United States
• Language: English

Related Subjects

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS
RIA
Transient Reactor Test
Zircaloy