Mirco Grosse

Flexible sample environment for high resolution neutron imaging at high temperatures in controlled atmosphere

High material penetration by neutrons allows for experiments using sophisticated sample environments providing complex conditions. Thus, neutron imaging holds potential for performing in situ nondestructive measurements on large samples or even full technological systems, which are not possible with any other technique. This paper presents a new sample environment for in situ high resolution neutron imaging experiments at temperatures from room temperature up to 1100 °C and/or using controllable flow of reactive atmospheres. The design also offers the possibility to directly combine imaging with diffraction measurements. Design, special features, and specification of the furnace are described. In addition, examples of experiments successfully performed at various neutron facilities with the furnace, as well as examples of possible applications are presented. This covers a broad field of research from fundamental to technological investigations of various types of materials and components.

In-situ investigation of hydrogen diffusion in Zircaloy-4 by means of neutron radiography

The hydrogen diffusion from the gas phase into Zircaloy-4 solid cylinders was investigated at various temperatures between 823 and 1473 K. Diffusion coefficients were fitted using axial hydrogen distributions for all temperatures investigated. The activation energy of the hydrogen diffusion was calculated for temperatures at which the bcc β phase is stable. The determined value of 48.3 kJ mol−1 K−1 is significantly higher than values published by other authors. At lower temperatures the hydrogen uptake is connected with a phase transformation from the hep α- to the bcc β-phase. The spatial phase distribution and the hydrogen concentrations in the two phases were analysed.

Quantitative analysis of hydrogen uptake, diffusion and distribution in nuclear fuel rod claddings made of zirconium alloys

Neutron News Volume 26 • Number 2 • 2015 31 The strong difference between the total neutron cross sections of hydrogen and zirconium provides the possibility of exand in-situ investigations of this system by means of neutron imaging [1–8]. The dependence of the total macroscopic neutron cross section on the hydrogen concentration can be determined by means of calibration specimens with known hydrogen concentration. As an example, Figure 1 shows such calibration measurements performed at the ANTARES facility at FRM-2, Garching, Germany. The calibration allows full quantitative analysis of the hydrogen concentrations [5]. Neutron imaging was applied to investigate the hydrogen uptake, diffusion and distribution in zirconium alloys used as material for nuclear fuel claddings. Neutron imaging experiments were performed using the facilities ICON (SINQ, PSI Villigen, Switzerland), ANTARES (FRM-2, Garching, Germany) and CONRAD (BER-II, Helmholtz Zentrum Berlin, Germany). In the following, examples for the investigation of the different processes will be given.

Parameters Influencing the Secondary Hydrogenation of Zr-Sn and Zr-Nb Alloys During Steam Oxidation Under LOCA and Severe Accident Conditions

The parameters influencing secondary hydrogen uptake can be divided into two groups: material properties and process parameters. The first group includes for instance the steam oxidation kinetics, the oxide morphology and the hydrogen diffusion through the oxide layer. The second group covers for instance the temperature, the total pressure, the gas flow type and rates, the cladding area and the filling of the rods. Together with a theoretical view on the influence of different parameters on the hydrogen uptake of zirconium alloys experimental results from separate-effect tests, large-scale QUENCH tests and in-situ neutron radiography investigations of the hydrogen uptake during steam oxidation will be presented. The hydrogen concentrations in specimens made from commonly used cladding materials were determined by quantitative analysis of neutron radiographs. Information obtained from ex- and in-situ steam oxidation experiments will be given. The presentation of the experimental results will be focused on the influence of oxidation time and temperature, of the oxide layer morphology, the sample geometry and of the gas flow rates on the hydrogen concentration of the remaining metal phases. Differences between Zr-Sn, Zr-Nb and Zr-Sn-Nb alloys will be discussed.Copyright

Influence of composition and heating schedules on compatibility of FeCrAl alloys with high-temperature steam

Abstract FeCrAl alloys are proposed and being intensively investigated as alternative accident tolerant fuel (ATF) cladding for nuclear fission application. Herein, the influence of major alloy elements (Cr and Al), reactive element effect and heating schedules on the oxidation behavior of FeCrAl alloys in steam up to 1500 °C was examined. In case of transient ramp tests, catastrophic oxidation, i.e. rapid and complete consumption of the alloy, occurred during temperature ramp up to above 1200 °C for specific alloys. The maximum compatible temperature of FeCrAl alloys in steam increases with raising Cr and Al content, decreasing heating rates during ramp period and doping of yttrium. Isothermal oxidation resulted in catastrophic oxidation at 1400 °C for all examined alloys. However, formation of a protective alumina scale at 1500 °C was ascertained despite partial melting. The occurrence of catastrophic oxidation seems to be controlled by dynamic competitive mechanisms between mass transfer of Al from the substrate and transport of oxidizing gas through the scale both toward the metal/oxide scale interface.

H2 PERMEATION BEHAVIOR OF Cr2AlC AND Ti2AlC MAX PHASE COATED ZIRCALOY-4 BY NEUTRON RADIOGRAPHY

Hydrogen uptake by nuclear fuel claddings during normal operation as well as loss of coolant during design basis and severe accidents beyond design basis has a high safety relevance because hydrogen degrade the mechanical properties of the zirconium alloys applied as cladding material. Currently, claddings with enhanced accident tolerance are under development. One group of such accident tolerant fuel (ATF) claddings are zirconium alloys with surface coatings reducing corrosion and high-temperature oxidation rate, as well as the chemical heat and hydrogen release during hypothetical accidents. The hydrogen permeation through the coating is an important parameter ensuring material safety. In this work, the hydrogen permeation of Ti2AlC and Cr2AlC MAX phase coatings on Zircaloy-4 is investigated by means of neutron radiography. Both coatings are robust hydrogen diffusion barriers that effectively suppress hydrogen permeation into the matrix.

High-temperature oxidation resistance and self-healing behavior of Cr2AlC MAX phase coating on Zircaloy-4

Zirconium-based alloys are currently utilized as fuel cladding and structural components in commercial light water reactors due to their low thermal neutron absorption cross section, good mechanical properties and reasonable corrosion resistance during operation conditions. One undesirable feature of zirconium-based alloy cladding is their extremely fast oxidation kinetics with high-temperature steam during loss of cooling accidents (LOCA). A considerable amount of heat and hydrogen gas is produced by the reaction of zirconium and steam. The claddings undergo severe degradation and hydrogen explosion can occur, followed by subsequent release of highly-radioactive fission products to the environment like during the nuclear accidents at the Fukushima Daiichi Nuclear Power Plant in 2011. One strategy to improve the accident tolerance of the state-of-the-art zirconium-based alloy fuel claddings is to coat the outer surface with an oxidation resistant coating. This solution promises the elimination of corrosion degradation during normal operation, as well as significant reduced oxidation kinetics with steam during off-normal conditions. Mn+1AXn(MAX) phases represent a family of ternary layered carbides or nitrides which possess a unique combination of the merits of both metals and ceramics. Alumina-forming MAX phase materials, like Ti2AlC and Cr2AlC, are being considered as protective coatings with respect to their excellent oxidation resistance up to 1400°C. In this study, Cr2AlC coatings have been deposited on Zircaloy-4 substrates by magnetron sputtering using elemental nano-multilayer thin films, and subsequent thermal annealing in argon. The total thickness of the coatings is around 6.5 μm and both coatings have a 500 nm Cr layer as bonding layer and diffusion barrier. One design of coatings also deposited a 1.5μm thick Cr capping layer to migrate potential fast dissolve of Al during normal operation. Crystallization of Cr2AlC MAX phase starts from 480°C by annealing in Ar and formation of phase-pure Cr2AlC MAX phase but with surface microcracks at 550°C is confirmed. Both coatings demonstrated high adherence, excellent oxidation resistance up to at least 1200°C and self-healing capability with growth of protective Al2O3 scale or of protective Al2O3 scale beneath Cr2O3 during high-temperature oxidation.