Burst behavior of nuclear grade FeCrAI and Zircaloy-2 fuel cladding under simulated cyclic dryout conditions

A novel experiment to simulate cyclic dryout in boiling water reactors has been developed to better understand the performance of nuclear grade FeCrAI cladding in a BWR during dryout conditions caused by an Anticipated Operational Occurrence or Anticipated Transient Without SCRAM - both of which are Design Basis Accidents. Internally pressurized C26 M FeCrAI alloy cladding and Zircaloy-2 cladding were subjected to rapid 300 degrees-650 degrees C thermal cycling in a steam environment; actual maximum temperatures were found to vary between materials but were always above 650 degrees C. In the range of 32-55 MPa hoop stress, Zircaloy-2 cladding burst within 1-16 cycles (about 100 s of dryout duration above 600 degrees C), while at 76 MPa hoop stress, C26 M cladding remained virtually undeformed after completing 54 cycles (over 1000 s of dryout duration above 600 degrees C). Higher temperature 300 degrees-700 degrees C and 300 degrees C-800 degrees C cycling experiments had to be performed to induce C26 M burst - failure occurred after 20 cycles in the former and during the first cycle in the latter. Zircaloy-2 and C26 M failure criteria were used to generate hoop stress specific dryout lifetimes. Overall, the simulated cyclic dryout experiments show that nuclear grade C26 M cladding has significantly enhanced survivability under dryout conditions relative to Zircaloy-2. (C) 2020 Elsevier B.V. All rights reserved.