Behavior of AglnCd Absorber Material in Zry/UO2 Fuel Rod Simulator Bundles Tested at High Temperatures in the CORA Facility

The CORA experiments carried out in an out-of-pile facility at the Kernforschungszentrum Karlsruhe (KfK), Federal Republic of Germany, are part of the "Severe Fuel Damage" (SFD) program. The experimental program is to provide information on the failure mechanisms of Light Water Reactor (LWR) fuel elements in a temperature range from 1200°C to 2000°C and in a few cases up to 2400°C. In the CORA experiments two different bundle configurations are tested: PWR (Pressurized Water Reactor) and BWR (Boiling Water Reactor) bundles. The PWR-type assemblies usually consist of 25 rods with 16 electrically heated fuel rod simulators and nine unheated rods (full-pellet and absorber rods). Bundle CORA-5 contained one Ag/ln/Cd - steel absorber rod whereas two absorber rods were used in CORA-12, CORA-15, and CORA-9. The larger bundle CORA-7 contained 5 absorber rods. CORA-12 was terminated by quenching with water from the bottom. In CORA-15 the heated and unheated rods were pressurized to achieve pronounced clad ballooning. Bundle CORA-9 was tested with a system pressure of 1.0 MPa instead of 0.22 MPa. The test bundles were subjected to temperature transients of a slow heatup rate in a steam environment. Thus, an accident sequence is simulated, which may develop from a small-break loss-of-coolant accident of a LWR. The transient phases of the tests were initiated with a temperature ramp rate of 1 K/s. The temperature escalation due to the exothermal zircaloy (Zry)-steam reaction started at about 1100°C, leading the bundles to maximum temperatures of approximately 2000°C. Rod destruction started with the failure of the absorber rod cladding at about 1200°C, i. e. about 250 K below the melting regime of steel. Penetration of the steel cladding was presumably caused by a eutectic interaction between steel and the zircaloy guide tube. The test bundles resulted in severe oxidation and partial melting of the cladding, fuel dissolution by Zry/UO 2 interaction, complete Inconel spacer destruction, and relocation of melts and fragments to lower elevations in the bundle, where extended blockages have formed. There was no difference in the behavior of the large bundle CORA-7 compared to the test bundles of regular size.