Steam and Air Oxidation Behavior of Nuclear Fuel Claddings at Severe Accident Conditions
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STEAM AND AIR OXIDATION BEHAVIOR OF NUCLEAR FUEL CLADDINGS AT SEVERE ACCIDENT CONDITIONS Mirco Große, Martin Steinbrück, and Juri Stuckert Karlsruhe Institute of Technology, Germany; Institute for Materials Research
ABSTRACT The oxidation behavior of zirconium alloys used as materials for nuclear fuel rod claddings is investigated in the temperature range between 973 and 1673 K in steam and air atmosphere. Parabolic kinetics was found for all materials, atmospheres and temperatures, at least at beginning of the reactions. The temperature dependence of the reaction rate is of Arrhenius type. The parameters of the Arrhenius functions are determined and given for steam oxidation. Due to the formation of a large amount of cracks an acceleration of the reactions can occur. Reasons of the crack formations are phase transformations in the oxide layer known as the breakaway effect and, in case of air atmosphere, local oxygen starvation conditions resulting in reactions with nitrogen. The paper gives a short overview of the relevant mechanisms and processes.
INTRODUCTION Hypothetical scenarios of severe accidents in nuclear reactors are studied experimentally in the QUENCH program performed at the Karlsruhe Institute of Technology (KIT, formerly FZK) [1]. In this program large scale tests with nuclear fuel rod bundle simulators were performed. Additionally, the large scale tests were supported by separate-effect tests to determine for instance isothermal and transient reaction kinetics or failure mechanisms. The accident scenarios start with a loss of the coolant. It results in an overheating of the reactor core. Reflood of the reactor core is the main measure to terminate such an accident. The combination of water and the overheated core results in steam oxidation of the nuclear fuel claddings made of zirconium alloys. Water molecules react with oxygen vacancies in the growing oxide layer. Also scenarios including air or pure nitrogen ingress into the reactor core can occur and were investigated. The oxidation behavior of zirconium alloys used as nuclear fuel rod cladding material at high temperatures in different atmospheres is widely studied (see [2, 3] and the references in these papers). However, a lot of questions are still open and some processes are not yet fully understood or can not be satisfyingly described in mechanistic or phenomenological models. This paper provides an overview of the isothermal reaction kinetics of several zirconium alloys currently used for fuel rod claddings in steam, oxygen, air and nitrogen in the temperature range between 973 and 1673 K.
MATERIALS AND MEASUREMENTS The investigations comprise the classical fuel rod cladding material Zircaloy-4 (Zry-4, Zr-Sn alloy), the Zr-Nb alloys M5™ and E110, the Zr-Nb-Sn alloy Zirlo™ and the composite material Dx/D4 consisting of a Zry-4 bulk and a corrosion protective layer D4 characterized by a
reduced tin and a higher iron content. Details about producers and main alloying elements are provided in Tab. 1. Segments with a length of
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