High-temperature oxidation and mutual interactions of materials during severe nuclear accidents

 M. Steinbrück, M. Große, J. Stuckert

Karlsruhe Institute of Technology, Institute for Applied Materials, Germany

ABSTRACT (Invited Talk)

During a nuclear accident with a loss of coolant, the reactor core steadily heats up due to the release of decay heat and reduced heat transfer to the remaining steam. The temperature rise extends up to the point where stability limits of some materials in the core structure are reached and complex chemical reactions are involved.

Oxidation of zirconium alloy cladding material becomes significant from temperatures of about 1000°C, causing mechanical degradation and a loss-of-barrier (against release of fission products) effect. Furthermore, this reaction is strongly exothermal, i.e. connected with release of heat comparable to and exceeding the residual nuclear power; and it is the main source of hydrogen during a nuclear accident jeopardizing the containment and reactor building (as seen during the Fukushima Daiichi accidents) and may be absorbed by metallic zirconium.

Nitrogen is used for inertization of boiling water reactor (BWR) containments and for pressurization of emergency cooling water systems and comes into play during air ingress scenarios. It strongly affects the oxidation kinetics by the formation of zirconium nitride and its re-oxidation. Due to the significantly different densities of ZrN and ZrO2, porous, non-protective oxide layers are formed over a wide temperature range. Depending on temperature, the oxidation of Zry in steam-nitrogen mixtures may be faster than the oxidation in steam by one order of magnitude.

The various core component materials are chemically unstable with respect to each other and eutectic interactions occur which lead to the formation of liquid phases in LWR fuel rod bundles at temperatures of approx. 1200°C already, i.e. significantly below the melting temperatures of the materials involved. Initial degradation occurs in the control rods with Ag-In-Cd alloy and boron carbide absorber materials. Whereas the low-temperature Ag-In-Cd alloy (used in most PWRs; melting temperature about 800°C) does not interact chemically with the enclosing stainless steel cladding, very rapid eutectic interactions between B4C (used in BWRs, recent PWRs, and VVERs; melting temperature 2450°C) and stainless steel as well as between stainless steel and zircaloy take place at about 1250°C. Failure of the Ag-In-Cd control rods is caused by high Cd vapor pressure and/or eutectic interaction between the surrounding steal and Zircaloy tubes. The resulting absorber melt may attack adjacent fuel rods and is an additional source of hydrogen and heat due to its rapid oxidation.

The paper discusses the materials interactions in the early phase of a severe nuclear accident and presents highlights of the corresponding research at KIT, including large-scale bundle experiments and separate-effects tests on the laboratory scale.