Chemical-physical behavior of light water reactor core components tested under severe reactor accident conditions in the CORA facility

Integral experiments with 2-m-long pressurized water reactor and boiling water reactor fuel rod bundle simulators containing a maximum of 57 rods (the CORA experimental program) as well as comprehensive single-effects investigations are examined. The physicochemical material behavior of light water reactor fuel elements up to ∼2700 K under flowing steam is described. Of particular importance is the determination of critical temperatures above which liquid phases form as a result of chemical interactions between the fuel element components and their influence on damage propagation. The results of the experiments show that low-temperature liquid phases form as early as ∼ 1300 K as a result of chemical interactions of INCONEL grid spacers with the Zircaloy cladding tube, of the absorber materials (Ag-In-Cd) with Zircaloy, and of boron carbide with stainless steel; however, extensive propagation of these interactions over large distances occurs only above 1550 K. Uranium oxide (UO 2 ) fuel can be liquefied (dissolved) by molten metallic Zircaloy, with the formation of a a U-Zr-O melt resulting in UO 2 relocation. This process can even take place below the melting point of Zircaloy (2040 K) if the melt, generated by chemical reactions with the various core components, contains metallic zirconium. Beyond the melting point of Zircaloy (≥2040 K), the metallic melt dissolves UO 2 more strongly; i.e., at a given time, more UO 2 is dissolved. In this case, UO 2 relocation occurs ∼ 1000 K below its melting point. The molten materials form coolant channel blockages (crusts) on solidification. In the CORA experimental facility, temperatures necessary to melt the remaining solid ceramic materials, up to ∼ 3150 K (according to the U-Zr-O phase diagram), were not attained. On the basis of the experimental results and thermodynamic considerations, three distinct temperature regimes can be defined where liquid phases that form in the reactor core give rise to substantial material relocations and different degrees of core damage. Quenching of an overheated fuel element with water from the bottom (simulating flooding of an uncovered reactor core) initially gives rise to further heating of the bundle components as a result of intensive oxidation of metallic constituents, which is associated with the formation of local melts and the additional generation of considerable amounts of hydrogen within a very short period of time.