# In Situ Neutron Radiography Investigations of Hydrogen Related Processes in Zirconium Alloys
## 1 Introduction  
Hydrogen in metals is an evergreen in material research. Even small amounts of hydrogen absorbed in a material can degrade the mechanical properties significantly. Several fracture mechanisms caused by hydrogen are well known. One example is the so-called fisheye fracture occurring in steels even with hydrogen concentrations in the order of magnitude of a few wt.ppm. Here, dissolved hydrogen (protons) moves to the crotches of grain boundaries, recombines with molecular hydrogen, and builds up an inner pressure, which can exceed the fracture strength even with subcritical external load. Hydrogen pickup degrades the mechanical properties of zirconium alloys used for nuclear fuel claddings too. Hydrogen is absorbed by water corrosion during operation or by high-temperature steam oxidation under accident conditions. The hydrogen concentration in spent fuel claddings has values between 50 and 80 wt.ppm depending on the chemical composition and the production paths of the tubes [1]. At high burn-up, concentrations of up to 650 wt.ppm can be reached. The solubility of hydrogen in Zircaloy-4 (Zr 1.3% Sn) at a reactor operation temperature of 300 \({}^{\circ}\)C is about 70 wt.ppm [2]. Most of the hydrogen is dissolved in the zirconium lattice during operational conditions up to a burn-up of about 50 GWd/MTU. For discharging of the fuel assemblies, the reactor is cooled down. The hydrogen solubility at room temperature is very low. Hydrides precipitate nearly exclusively with orientation on a circumferential direction because the outer pressure is higher than the inner pressure in the fuel rods (see Figure 1). After storage in a spent fuel pool at temperatures of about 60 \({}^{\circ}\)C, the fuel rods heat up to temperatures between 350 and 400 \({}^{\circ}\)C at the beginning of the dry storage due to the remaining decay heat. Most of the hydrides become dissolved again (hydrogen solubility at 400 \({}^{\circ}\)C is about 200 wt.ppm). The decrease in decay heat results in a very slow cool-down of the spent fuel rods, and with it to a reprecipitation of the hydrides. Becausethe inner pressure in the rods is higher than the outer pressure under dry storage conditions, the hydrides can be oriented in a radial direction. This orientation degrades the mechanical strength much stronger than hydrides oriented in a circumferential direction. In order to develop models describing these processes over a time scale of 50 to 100 years, the influence of system parameters such as temperature, texture, grain size, chemical composition, and elastic strain on the hydrogen diffusivity and solubility has to be determined accurately. During design basis large break loss of coolant accidents (LB-LOCAs), band-shaped hydrogen enrichments with concentrations of several hundred wt.ppm can be formed due to the so-called secondary hydrogen uptake via the inner surface of the cladding tubes after failure. Hydrogen concentrations of up to more than 3500 wt.ppm were detected in cladding tubes applied in large-scale severe accident simulation tests. Figure 2 shows the distribution of these high hydrogen concentrations. It results in an embrittlement of the materials as it was shown in [3]. The sample fails in the tensile test at the hydrogen enrichments if the local hydrogen concentration exceeds 1500 wt.ppm. This material degradation provides the risk of brittle thermo-shock fracture during emergency cooling, terminating the accident. If this happens, fuel and fission products could be released and placed on the bottom of the reactor vessel, resulting in coolability issues. Therefore, one focus of the actual nuclear reactor safety research is the investigation of the concentration, shape, and dimension of the hydrogen enrichments after several accident scenarios. Neutron imaging is a powerful tool for the investigation of hydrogen in zirconium alloys because the macroscopic total neutron cross section of hydrides can be up to 2500% compared with one of the typical zirconium alloys applied as cladding material.