Investigations of the hydrogen diffusion and distribution in Zirconium by means of Neutron Imaging
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M. Grosse
, J. R. Santisteban , J. Bertsch , B. Schillinger , A. Kaestner , M. R. Daymond und N. Kardjilov
Abstract
Absorbed hydrogen degrades the mechanical properties of zirconium alloys used for nuclear fuel claddings. Not only the total amount of hydrogen absorbed in the cladding tube but also the zirconium hydride orientation and its distribution influence the toughness of the material. For instance, the so-called delayed hydride cracking is caused by the diffusive re-distribution of hydrogen into the dilative elastic strain field ahead of crack tips. The paper presents in-situ and ex-situ neutron imaging investigations of hydrogen uptake, diffusion and distribution in zirconium alloys used for claddings. An overview about results of in-situ experiments studying the hydrogen uptake in strained Zircaloy-4, as well as ex-situ investigations of the diffusion of hydrogen in cold rolled Zircaloy-2 and Zr-2.5 % Nb alloy depending on temperature, rolling direction and thermal treatment and of the hydrogen re-distribution in the β-phase of Zircaloy-4 during a Three-Point-Bending-Test at 600 °C are presented.
Kurzfassung
Als Material für Brennstabhüllrohre in Leichtwassereaktoren werden Zirkonium-Legierungen verwendet. Wenn diese Wasserstoff aufnehmen, verschlechtern sich ihre mechanischen Eigenschaften. Dabei wird die Zähigkeit der Materialien nicht nur durch die Menge des absorbierten Wasserstoffs, sondern auch von der Orientierung und Verteilung der Hydride beeinflusst. Zum Beispiel ist der sogenannte verzögerte Wasserstoffbruch von der diffusionsgesteuerten Umverteilung des Wasserstoffs zum Gebiet vor einer Rissspitze verursacht, welches elastisch gedehnt ist. Dieser Artikel präsentiert Ergebnisse von in-situ Neutronenradiographie-Studien der Wasserstoffaufnahme in gedehnten Zircaloy-4-Proben, und ex-situ Untersuchungen zur Wasserstoffdiffusion in kaltgewalzten Zircaloy-2 und Zr-2.5 %Nb-Legierungen in Abhängigkeit von Temperatur, Walzrichtung und Wärmebehandlung sowie zur Umverteilung des Wasserstoffs in der β-Phase von Zircaloy-4 während eines Drei-Punkt-Biege-Test bei 600 °C.
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© 2018, Carl Hanser Verlag, München
Artikel in diesem Heft
- Contents/Inhalt
- Contents
- Editorial
- Safety of extended dry storage of spent nuclear fuel – GRS workshop 2018
- Technical Contributions/Fachbeiträge
- CIEMAT response to challenges on fuel safety research during dry storage
- Research activities at GRS on fuel rod behaviour during extended dry storage
- Open questions on the road to reliable predictions of cladding integrity
- Considerations on spent fuel behavior for transport after extended storage
- Investigations of the hydrogen diffusion and distribution in Zirconium by means of Neutron Imaging
- Effect of Zirconium Hydrides on the mechanical behavior of cladding
- Response of irradiated nuclear fuel rods to quasi-static and dynamic loads
- Investigations on potential methods for the long-term monitoring of the state of fuel elements in dry storage casks
Artikel in diesem Heft
- Contents/Inhalt
- Contents
- Editorial
- Safety of extended dry storage of spent nuclear fuel – GRS workshop 2018
- Technical Contributions/Fachbeiträge
- CIEMAT response to challenges on fuel safety research during dry storage
- Research activities at GRS on fuel rod behaviour during extended dry storage
- Open questions on the road to reliable predictions of cladding integrity
- Considerations on spent fuel behavior for transport after extended storage
- Investigations of the hydrogen diffusion and distribution in Zirconium by means of Neutron Imaging
- Effect of Zirconium Hydrides on the mechanical behavior of cladding
- Response of irradiated nuclear fuel rods to quasi-static and dynamic loads
- Investigations on potential methods for the long-term monitoring of the state of fuel elements in dry storage casks