4^th SCIENTIFIC & TECHNOLOGICAL CONFERENCE

DIAGNOSTICS OF MATERIALS AND INDUSTRIAL COMPONENTS

31.05 - 2.06 2012, GDANSK UNIVERSITY OF TECHNOLOGY

ABSTRACT

Materials behavior during the early phase of a severe nuclear accident

Martin Steinbrück

Karlsruhe Institute of Technology, Institute of Applied Materials IAM-AWP, GERMANY

martin.steinbrueck@kit.edu, phone: +49-721-608-22517

After loss of coolant in a nuclear power plant (here only light water reactors, LWRs, are discussed) temperatures in the core rise due the residual decay heat even the reactor was successfully shut down. Starting from about 1000°C the oxidation of the zirconium alloy (Zry) claddings, enclosing the UO₂ fuel pellets, becomes significant causing mechanical degradation of the cladding rods as well as release of hydrogen and chemical heat. For example, the hydrogen produced by the zirconium-steam reaction caused the detonations of the reactor buildings during the Fukushima Daiichi accidents. At temperatures above ca. 1500°C the heat produced by this reaction is in the range of and even higher than the decay heat and hence strongly influences the progress of the accident.

The oxidation kinetics of currently applied cladding alloys at temperatures 600-1600°C in various atmospheres was extensively investigated at Karlsruhe Institute of Technology (KIT) during the last decade. Generally, parabolic rate equations are applied in severe accident codes which are determined by the growth of a protective superficial oxide scale. However, at temperatures below 1100°C a transition to accelerated, more or less linear kinetics was found for most of the alloys after critical oxide scale thicknesses were exceeded. This transition is caused by the so-called breakaway, i.e. the formation of non-protective oxide layers.

Nitrogen is used for inertization of boiling water reactor (BWR) containment 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 can be by one order of magnitude faster than the oxidation in only steam.

Absorber materials may have also strong impact on core degradation and fission product behavior. Boron carbide (B₄C) is widely used as neutron absorbing control rod material in Western boiling water reactors and recent pressurized water reactors (PWR) as well as in Russian VVERs. It was also applied in all units of the Fukushima Dai-ichi nuclear power plant. Usually it is enclosed by stainless steel (SS) in the form of cladding tubes or blades. Although the melting temperature of B₄C is at about 2450°C, it initiates local, but significant melt formation in the core at temperatures around 1250°C due to eutectic interactions with the surrounding SS and Zry structures. The B₄C containing melt relocates and hence transports material and energy to lower parts of the fuel bundle. It is chemically aggressive and may attack other structure materials. Furthermore, boron carbide and absorber melt are oxidized by steam very rapidly and thus contribute to the hydrogen source term in the early phase of a severe accident.

Silver-Indium-Cadmium (SIC) alloy is applied in PWRs. It has a low melting temperature (800°C), but the enclosing SS cladding is chemically stable against the alloy. Failure of SIC control rods is observed beyond 1200°C due to eutectic interaction SS with Zry and/or mechanical break of the SS cladding.

The paper presents highlights of the corresponding research at KIT including large-scale bundle experiments and separate-effects tests in laboratory scale.