Mechanical properties of neutron-irradiated model and commercial FeCrAl alloys
Source abstract
Abstract
Abstract The development and understanding of the mechanical properties of neutron-irradiated FeCrAl alloys is increasingly a critical need as these alloys continue to become more mature for nuclear reactor applications. This study focuses on the mechanical properties of model FeCrAl alloys and of a commercial FeCrAl alloy neutron-irradiated to up to 13.8 displacements per atom (dpa) at irradiation temperatures between 320 and 382 °C. Tensile tests were completed at room temperature and at 320 °C, and a subset of fractured tensile specimens was examined by scanning electron microscopy. Results showed typical radiation hardening and embrittlement indicative of high chromium ferritic alloys with strong chromium composition dependencies at lower doses. At and above 7.0 dpa, the mechanical properties saturated for both the commercial and model FeCrAl alloys, although brittle cleavage fracture was observed at the highest dose in the model FeCrAl alloy with the highest chromium content (18 wt %). The results suggest the composition and microstructure of FeCrAl alloys plays a critical role in the mechanical response of FeCrAl alloys irradiated near temperatures relevant to light water reactors.
Source: openaire · Quality: candidate unverified · Source record ↗
Research context
OpenAlex topics: Nuclear Materials and Properties · Fusion materials and technologies · Nuclear reactor physics and engineering
SciVX discovery tags are terminology-based. OpenAlex topics are source-provided classifications.
References (39)
Show source-provided references
1. 1965
2. 1966
3. 1966
4. Kangilaski · 1967
5. ORNL · 1996
6. Preparation and properties of some ODS Fe-Cr-Al alloys10.1007/BF02646711 ↗
7. Advanced oxidation-resistant iron-based alloys for LWR fuel cladding10.1016/j.jnucmat.2013.06.041 ↗
8. Accident tolerant fuels for LWRs: a perspective10.1016/j.jnucmat.2013.12.005 ↗
9. Material selection for accident tolerant fuel cladding
10. High temperature oxidation of fuel cladding candidate materials in steam–hydrogen environments10.1016/j.jnucmat.2013.05.047 ↗
11. The effect of steam on the high temperature oxidation behavior of alumina-forming alloys10.1179/0960340914Z.00000000058 ↗
12. Development and property evaluation of nuclear grade wrought FeCrAl fuel cladding for light water reactors10.1016/j.jnucmat.2015.10.019 ↗
13. Radiation tolerance of neutron-irradiated model Fe–Cr–Al alloys10.1016/j.jnucmat.2015.06.023 ↗
14. Stress corrosion cracking of ferritic-maretensitic steels in simulated boiling water reactor environment10.5006/1806 ↗
15. Yamamoto · 2015
16. Irradiation-enhanced α′ precipitation in model FeCrAl alloys10.1016/j.scriptamat.2016.02.002 ↗
17. A combined APT and SANS investigation of alpha prime phase precipitation in neutron-irradiated model FeCrAl alloys10.1016/j.actamat.2017.02.077 ↗
18. Method for analyzing passive silicon carbide thermometry with a continuous dilatometer to determine irradiation temperature10.1016/j.nimb.2016.01.005 ↗
19. Void-swelling in irons and ferritic steels10.1016/0022-3115(79)90123-5 ↗
20. Behavior of different austenitic stainless steels, conventional, reduced activation (RA) and ODS chromium-rich ferritic-martensitic steels under neutron irradiation at 325°C in PWR environment
21. Role of scale factor during tensile testing of small specimens
22. Effect of specimen thickness on the tensile deformation properties of SA508 Cl.3 reactor pressure vessel steel
23. Observation of serrated flow in APMTTM steel10.1016/j.matlet.2015.07.072 ↗
24. Tensile properties of accident-tolerant aluminum-bearing ferritic steels10.1016/j.anucene.2016.09.018 ↗
25. Mechanical performance of ferritic martensitic steels for high dose applications in advanced nuclear reactors10.1007/s11661-012-1565-y ↗
26. Application of a three-feature dispersed-barrier hardening model to neutron-irradiated Fe–Cr model alloys10.1016/j.jnucmat.2014.01.024 ↗
27. Neutron irradiation effects in Fe and Fe-Cr at 300 °C10.1016/j.actamat.2016.03.060 ↗
28. Small angle neutron scattering study of irradiated martensitic steels: relation between microstructural evolution and hardening
29. Mechanical properties of spinodally decomposed Fe-30 wt% Cr alloys: yield strength and aging embrittlement10.1016/0001-6160(86)90130-6 ↗
30. On the origin of hardening in aged FeCr alloys: relation between unmixing and mechanical properties
31. Microstructure and fracture behavior of F82H steel under different irradiation and tensile test conditions10.1016/j.jnucmat.2015.09.031 ↗
32. Void-swelling in irons and ferritic steels: I. mechanisms of swelling suppression10.1016/0022-3115(79)90122-3 ↗
33. Interstitial-impurity radiation-defect interactions in ferritic steels10.1007/BF03258766 ↗
34. Impurity-defect interactions and radiation hardening and embrittlement in BCC metals10.1007/BF02670157 ↗
35. Dynamic strain aging in SA508-class 3 pressure vessel steel10.1016/0308-0161(95)93969-C ↗
36. Influences of the dynamic strain aging on the J–R fracture characteristics of the ferritic steels for reactor coolant piping system10.1016/j.nucengdes.2003.06.014 ↗
37. Development of microstructure and irradiation hardening of Zircaloy during low dose neutron irradiation at nominally 358°C10.1016/j.jnucmat.2011.07.006 ↗
38. Alloy selection for accident tolerant fuel cladding in commercial light water reactors
39. Uniform corrosion of FeCrAl alloys in LWR coolant environments10.1016/j.jnucmat.2016.06.047 ↗