Post-Irradiation Annealing of Dislocation Microstructure and Radiation-Induced Segregation in Proton-Irradiated Stainless Steels

1998 ◽  
Vol 540 ◽  
Author(s):  
J.T. Busby ◽  
G.S. Was ◽  
E.A. Kenik
Keyword(s):  
Stainless Steel ◽  
Dislocation Loop ◽  
Boundary Segregation ◽  
304L Stainless Steel ◽  
Dislocation Loops ◽  
Appreciable Change ◽  
Post Irradiation ◽  
Ultra High Purity ◽  
Before And After ◽  
Radiation Induced

AbstractBoth microstructural effects (radiation hardening) and microcompositional effects (radiationinduced segregation (RIS)) have been identified as potential contributors to irradiation-assisted stress corrosion cracking (IASCC). However, the importance of each in IASCC is unclear. In an effort to isolate their roles, post-irradiation annealing of proton-irradiated 304L stainless steel samples has been studied. Model simulations of post-irradiation annealing at intermediate temperatures (400-650°C) indicate that microstructural features such as dislocation loops are removed faster than RIS. Simulations also predict that there exist time-temperature combinations that will significantly reduce the dislocation loop population while leaving the grain boundary segregation essentially unaffected. Ultra-high purity (UHP) 304L stainless steel samples have been irradiated with 3.2 MeV protons at 360°C to 1.0 dpa and then annealed in-situ as thinned TEM disks in the temperature range of 500°C- 625°C for times between 20 and 60 minutes. RIS and dislocation loops were characterized before and after annealing in the same areas of each specimen. Post-irradiation anneals at 500TC for 45 minutes or 60 minutes resulted in no appreciable change in dislocation loop population or RIS. Annealing at 600°C for 20 minutes., or 40 minutes and at 625°C for 40 minutes, resulted in decreasing dislocation loop densities and increasing loop size with increasing annealing time or temperature, while the amount of RIS did not change significantly from the pre-annealed condition.

Nucleation and amorphization of radiation-produced phases in a modified austenitic stainless steel during Ni-ion irradiation

1988 ◽  
Vol 3 (5) ◽  
pp. 840-844 ◽  
Author(s):  
E. H. Lee ◽  
E. A. Kenik
Keyword(s):  
Stainless Steel ◽  
Austenitic Stainless Steel ◽  
Ion Irradiation ◽  
Amorphous State ◽  
Dislocation Loops ◽  
Nickel Ion ◽  
Carbide Phases ◽  
Direct Formation ◽  
Radiation Induced ◽  
Lower Temperature

The nucleation and amorphization of radiation-induced (G) and radiation-enhanced (η) phases in a silicon- and titanium-modified austenitic stainless steel have been studied under nickel-ion irradiation. These silicon- and nickel-enriched phases form under high-temperature (950 K) irradiation as the result of radiation-induced segregation to radiation-produced interstitial dislocation loops. Availability of carbon promotes the formation of η phase relative to G phase. Under lower temperature (450 K) irradiation, G and η phases are amorphized without significant change in composition of metallic elements. Two carbide phases (MC, M23C6) remain crystalline for the same irradiation conditions. The amorphization of the silicides may result from (1) radiation damage increasing their free energy above that of the amorphous state or (2) direct formation of the amorphous phase in the damage cascade.

Microstructure and Post-Irradiation Annealing Behavior of 20% Cold-Worked 316 Stainless Steel

2000 ◽  
Vol 650 ◽  
Author(s):  
J. I. Cole ◽  
T. R. Allen ◽  
H. Kusanagi ◽  
K. Dohi ◽  
J. Ohta
Keyword(s):  
Stainless Steel ◽  
Cold Work ◽  
Microstructural Development ◽  
Dislocation Loops ◽  
Annealing Behavior ◽  
Post Irradiation ◽  
316 Stainless Steel ◽  
Dose Rates ◽  
Cold Worked

ABSTRACTMicrostructural examination and in situ post-irradiation annealing studies were carried out on 20% cold-worked 316 stainless steel (SS) hexagonal duct material following irradiation in the reflector region of the EBR-II reactor. Stainless steel hexagonal ducts were used to house reactor subassemblies and provide a valuable source of information on irradiation behavior of reactor structural materials at lower dose-rates (on the order of 10-8 dpa/sec) than previously examined. The microstructural development of samples irradiated to doses of 1, 20 and 30 dpa is examined, while the post-irradiation annealing behavior of a sample irradiated to 20 dpa is described. Annealing studies were performed at 370 and 500°C to examine the kinetics of radiation damage recovery as a function of annealing temperature. The initial (pre-annealed) microstructures consists of a substantial density of irradiation induced chromium-rich M23C6 and M6C carbides which form both on the grain boundaries and within the grain interiors. Recovery of the cold- work is evident in the 1 dpa sample while samples irradiated to 20 and 30 dpa possess dense populations of voids and dislocation structures consisting of networks of line dislocations and faulted dislocation loops. Results indicate that post-irradiation annealing of the samples at 370°C for 1 hour has little effect on the microstructure, while further annealing at 500°C for 1 hour results in void shrinkage, the formation of small cavities, and a reduction in the dislocation loop and network density.

Experimental Investigation of Hydrogen Effect on Mechanical Properties of 304L Stainless Steel

2008 ◽  
Author(s):  
Yong Jiang ◽  
Jianming Gong ◽  
Jianqun Tang ◽  
Luyang Geng
Keyword(s):  
Mechanical Properties ◽  
Stainless Steel ◽  
304L Stainless Steel ◽  
Hydrogen Effect ◽  
Reliable Operation ◽  
Hydrogen Charging ◽  
Before And After ◽  
Hydrogenation Reactor ◽  
Purified Terephthalic Acid ◽  
Longitudinal Cracks

Hydrogenation reactor is one of the key equipments in the production process of purified terephthalic acid (PTA). The safe and reliable operation of hydrogenation reactor decides the safety of PTA installation. Service experience has shown that some longitudinal cracks occurred on the internal wall of the 304L stainless steel (SS) hydrogen nozzle. In this paper, a series of electrochemical hydrogen charging tests were carried out for 304L stainless steel at the ambient temperature. The comparison on mechanical properties was investigated before and after hydrogen charging in order to investigate the effect of mechanical properties of 304L SS. Fractography was analyzed by using scanning electron microscope (SEM). The results showed that the mechanical strength of 304L SS was not basically lost after hydrogen charging, but the ductility reduced in a certain extent. The fractography also showed the hydrogen embrittlement effect on the fracture.

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