Mechanical and microstructural analysis of an EM10 wrapper tube after neutron irradiation in Phénix
Source abstract
Abstract
Abstract The 9Cr-1Mo tempered martensitic steel was irradiated with fast neutrons in the French Phenix reactor. Mechanical tests, density measurement as well as microstructural investigations were carried out following irradiation up to 155 dpa in the temperature range 386°C – 525°C. Tensile, elongation and impact properties were only slightly modified by irradiation. Microstructural investigations revealed the presence of cavities, however the resulting swelling was low (0.1%). In most cases, the dislocation microstructure was arranged in a network. Nevertheless, some large loops were observed after irradiation at 398°C to 112 dpa. Overall, the EM10 steel showed excellent radiation resistance, which confirmed that this steel is an attractive wrapper candidate material for future sodium-cooled GEN IV fast reactors.
Source: openaire · Quality: candidate unverified · Source record ↗
Research context
OpenAlex topics: Fusion materials and technologies · Nuclear Materials and Properties · Nuclear reactor physics and engineering
SciVX discovery tags are terminology-based. OpenAlex topics are source-provided classifications.
References (25)
Show source-provided references
1. Fazio · Proc. Int. Conf. Fast React. Relat. Fuel Cycles Safe Technol. Sustain. Scenar. · 2013 · 1 · 487
2. Microstructural evolution of ferritic-martensitic steels irradiated in the fast breeder reactor Phénix10.1016/0022-3115(93)90080-I ↗
3. Comparison of swelling and irradiation creep behavior of fcc-austenitic and bcc-ferritic/martensitic alloys at high neutron exposure10.1016/S0022-3115(99)00225-1 ↗
4. Pre- and post-irradiation mechanical properties of ferritic-martensitic steels for fusion applications: EM10 base metal and EM10/EM10 welds10.1016/0022-3115(94)90128-7 ↗
5. Tensile and impact properties of 9Cr tempered martensitic steels and ODS-FeCr alloys irradiated in a fast reactor at 325°C up to 78 dpa10.1016/j.jnucmat.2010.12.203 ↗
6. Séran · 1992
7. Mechanical properties of 9Cr martensitic steels and ODS-FeCr alloys after neutron irradiation at 325°C up to 42dpa10.1016/j.jnucmat.2007.03.166 ↗
8. Mechanical behavior of reduced-activation and conventional martensitic steels after neutron irradiation in the range 250–450°C10.1016/S0022-3115(00)00076-3 ↗
9. Post-Irradiation Tensile Behavior and Residual Activity of Several Ferritic/Martensitic and Austenitic Steels Irradiated in Osiris Reactor at 325°C up to 9 dpa10.1520/JAI12354 ↗
10. Brachet · Int. Symp. Mater. Ageing Compon. Life Ext. · 1995 · 75
11. Wall · 1983 · 260/2
12. The effect of phosphorus on the ductility of 9Cr-1Mo steels10.1016/0956-7151(90)90022-9 ↗
13. Alamo · Miner. Met. Mater. Soc. · 1996 · 121
14. Gilbon · 1990
15. Impact properties of irradiated HT9 from the fuel duct of FFTF10.1016/j.jnucmat.2011.11.059 ↗
16. The EU programme for modelling radiation effects in fusion reactor materials: An overview of recent advances and future goals10.1016/j.jnucmat.2008.12.301 ↗
17. Dynamic observations of heavy-ion damage in Fe and Fe–Cr alloys10.1016/j.jnucmat.2009.02.003 ↗
18. Klueh · 2001
19. Neutron-induced swelling and embrittlement of pure iron and pure nickel irradiated in the BN-350 and BOR-60 fast reactors10.1016/j.jnucmat.2008.01.015 ↗
20. High swelling rates observed in neutron-irradiated V-Cr and V-Si binary alloys10.1016/0022-3115(92)90613-P ↗
21. Use of double and triple-ion irradiation to study the influence of high levels of helium and hydrogen on void swelling of 8–12% Cr ferritic-martensitic steels10.1016/j.jnucmat.2015.07.012 ↗
22. L. Belin, C. Rabourg, G.M. Decroix, X. Averty, 2003.
23. Irradiation creep and swelling of the US fusion heats of HT9 and 9Cr-1Mo to 208 dpa at ~ 400°C10.1016/0022-3115(94)90131-7 ↗
24. SANS and TEM of ferritic–martensitic steel T91 irradiated in FFTF up to 184dpa at 413°C10.1016/j.jnucmat.2013.04.025 ↗
25. Microstructural examination of commercial ferritic alloys at 200 dpa10.1016/S0022-3115(96)00222-X ↗