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Journal articles on the topic 'Irradiated stainless steel'

Author: Grafiati

Published: 4 June 2021

Last updated: 1 February 2022

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1

Kanne, W. R., G. T. Chandler, D. Z. Nelson, and E. A. Franco-Ferreira. "Welding irradiated stainless steel." Journal of Nuclear Materials 225 (August 1995): 69–75. http://dx.doi.org/10.1016/0022-3115(94)00439-0.

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2

Wang, C. A., M. L. Grossbeck, N. B. Potluri, and B. A. Chin. "Welding of irradiated stainless steel." Journal of Nuclear Materials 233-237 (October 1996): 213–17. http://dx.doi.org/10.1016/s0022-3115(96)00203-6.

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3

de Visser-Týnová, Eva, Stephen W. Swanton, Stephen J. Williams, Marcel P. Stijkel, Alison J. Walker, and Robert L. Otlet. "14C release from irradiated stainless steel." Radiocarbon 60, no. 6 (November 22, 2018): 1671–81. http://dx.doi.org/10.1017/rdc.2018.134.

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ABSTRACTRadiocarbon (14C or carbon-14, half-life 5730 yr) is a key radionuclide in the assessment of the safety of a geological disposal facility (GDF) for radioactive waste. In particular, the radiological impact of gaseous carbon-14 bearing species has been recognized as a potential issue. Irradiated steels are one of the main sources of carbon-14 in the United Kingdom’s radioactive waste inventory. However, there is considerable uncertainty about the chemical form(s) in which the carbon-14 will be released. The objective of the work was to measure the rate and speciation of carbon-14 release from irradiated 316L(N) stainless steel on leaching under high-pH anoxic conditions, representative of a cement-based near field for low-heat generating wastes. Periodic measurements of carbon-14 releases to both the gas phase and to solution were made in duplicate experiments over a period of up to 417 days. An initial fast release of carbon-14 from the surface of the steel is observed during the first week of leaching, followed by a drop in the rate of release at longer times. Carbon-14 is released primarily to the solution phase with differing fractions released to the gas phase in the two experiments: about 1% of the total release in one and 6% in the other. The predominant dissolved carbon-14 releases are in inorganic form (as 14C-carbonate) but also include organic species. The predominant gas-phase species are hydrocarbons with a smaller fraction of 14CO (which may include some volatile oxygen-containing carbon-species). The experiments are continuing, with final sampling and termination planned after leaching for a total of two years.

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4

Mills, W. J. "Fracture Toughness of Irradiated Stainless Steel Alloys." Nuclear Technology 82, no. 3 (September 1988): 290–303. http://dx.doi.org/10.13182/nt88-a34130.

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5

Brimhall, J. L., J. I. Cole, and S. M. Bruemmer. "Deformation microstructures in ion-irradiated stainless steel." Scripta Metallurgica et Materialia 30, no. 11 (June 1994): 1473–78. http://dx.doi.org/10.1016/0956-716x(94)90248-8.

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6

Kenik, E. A., J. T. Busby, M. K. Miller, A. M. Thuvander, and G. Was. "Grain Boundary Segregation and Irradiation-Assisted Stress Corrosion Cracking of Stainless Steels." Microscopy and Microanalysis 5, S2 (August 1999): 760–61. http://dx.doi.org/10.1017/s1431927600017128.

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Irradiation-assisted stress corrosion cracking (IASCC) of irradiated austenitic stainless steels has been attributed to both microchemical (radiation-induced segregation (RIS)) and microstructural (radiation hardening) effects. The flux of radiation-induced point defects to grain boundaries results in the depletion of Cr and Mo and the enrichment of Ni, Si, and P at the boundaries. Similar to the association of stress corrosion cracking with the depletion of Cr and Mo in thermally sensitized stainless steels, IASCC is attributed in part to similar depletion by RIS. However, in specific heats of irradiated stainless steel, “W-shaped” Cr profiles have been observed with localized enrichment of Cr, Mo and P at grain boundaries. It has been show that such profiles arise from pre-existing segregation associated with intermediate rate cooling from elevated temperatures. However, the exact mechanism responsible for the pre-existing segregation has not been identified.Two commercial heats of stainless steel (304CP and 316CP) were forced air cooled from elevated temperatures (∽1100°C) to produce pre-existing segregation.

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7

Chen, Y., and E. Marquis. "Microstructural Characterization of an Irradiated 304 Stainless Steel." Microscopy and Microanalysis 19, S2 (August 2013): 1746–47. http://dx.doi.org/10.1017/s1431927613010726.

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8

Furutani, Gen, Nobuo Nakajima, Takao Konishi, and Mitsuhiro Kodama. "Stress corrosion cracking on irradiated 316 stainless steel." Journal of Nuclear Materials 288, no. 2-3 (February 2001): 179–86. http://dx.doi.org/10.1016/s0022-3115(00)00704-2.

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9

Jin, Hyung-Ha, Eunsol Ko, and Junhyun Kwon. "Microstructural Characterization of Hydrogen Irradiated Austenitic Stainless Steel." Microscopy and Microanalysis 21, S3 (August 2015): 1001–2. http://dx.doi.org/10.1017/s1431927615005802.

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10

Watanabe, K., S. Jitsukawa, S. Hamada, T. Kodaira, and A. Hishinuma. "Weldability of neutron-irradiated type 316 stainless steel." Fusion Engineering and Design 31, no. 1 (April 1996): 9–15. http://dx.doi.org/10.1016/0920-3796(95)00424-6.

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11

Ford, I. J. "Intergranular fracture of fast reactor irradiated stainless steel." Acta Metallurgica et Materialia 40, no. 1 (January 1992): 113–22. http://dx.doi.org/10.1016/0956-7151(92)90204-r.

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12

Ford, I. J. "Transgranular fracture of Fast Reactor irradiated stainless steel." Journal of Nuclear Materials 182 (May 1991): 52–59. http://dx.doi.org/10.1016/0022-3115(91)90414-3.

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13

Kenik, E. A., and M. G. Burke. "Segregation in a neutron-irradiated Type 316 stainless steel." Proceedings, annual meeting, Electron Microscopy Society of America 50, no. 2 (August 1992): 1218–19. http://dx.doi.org/10.1017/s0424820100130729.

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Radiation-induced segregation (RIS) and associated irradiation-assisted stress corrosion cracking (IASCC) of austenitic alloys may be a major factor in limiting component lifetimes in water-cooled nuclear reactors. There are some similarities between radiation-induced sensitization/IASCC and thermally-induced sensitization/intergranular stress corrosion cracking. Both processes are associated with chromium depletion at grain boundaries. Segregation to boundaries in a neutron irradiated type 316 stainless steel has been investigated with both energy-dispersive X-ray spectrometry (EDXS) and parallel detection electron energy loss spectrometry (PEELS).All specimens were from the same heat of cold-worked type 316 stainless steel. Both unirradiated control material and material irradiated at ∼300°C to a range of fluences 0.3 - 5 × 1026 neutrons/m2 (E>0.1 MeV) were available. The mass of irradiated material was minimized by mechanically polishing 3-mm-diam. disks to ∼75 μm thickness prior to electropolishing. However, the specific radioactivity of the specimens, which increased with neutron fluence, limited the application of EDXS to the unirradiated and the lowest fluence irradiated materials.

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14

Hojna, Anna, Jan Michalicka, and Ondrej Srba. "Fracture of Irradiated Austenitic Stainless Steel with Special Reference to Irradiation Assisted Stress Corrosion Cracking." Key Engineering Materials 592-593 (November 2013): 569–72. http://dx.doi.org/10.4028/www.scientific.net/kem.592-593.569.

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This paper deals with fracture of neutron irradiated austenitic Ti-stabilized stainless steel 08Ch18N10T. The steel had been tested in air and in water environment (320°C) using several tests representing different stress strain conditions for crack initiation and growth; Slow Strain Rate and Crack Growth Rate tests were performed in the water. Without irradiation the steel did not suffer from stress corrosion cracking in the water, but on irradiated specimens appeared areas of intergranular fracture mixed with transgranular cleavage-like facets and secondary cracks typical for IASCC phenomenon. The differences between fracture of irradiated and non-irradiated specimens in air and in water are documented and discussed.

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15

USAMI, Hiroshi, and Yuuko NAGATORO. "Surface analysis of stainless steel irradiated by synchotron radiation." SHINKU 34, no. 3 (1991): 174–77. http://dx.doi.org/10.3131/jvsj.34.174.

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16

NOGAMI, Shuhei, Yuki SATO, and Akira HASEGAWA. "Fatigue Crack Initiation in Proton-Irradiated Austenitic Stainless Steel." Journal of Nuclear Science and Technology 48, no. 9 (September 2011): 1265–71. http://dx.doi.org/10.1080/18811248.2011.9711815.

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17

Zinkle, S. J., and R. L. Sindelar. "Defect microstructures in neutron-irradiated copper and stainless steel." Journal of Nuclear Materials 155-157 (July 1988): 1196–200. http://dx.doi.org/10.1016/0022-3115(88)90495-3.

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18

Nakabayashi, Ryo, and Tomonari Fujita. "Identification of Chemical Form of Carbon Released from SUS304 and SUS316 in Alkaline Solution under Low-oxygen Condition." MRS Advances 2, no. 11 (December 23, 2016): 597–602. http://dx.doi.org/10.1557/adv.2016.643.

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ABSTRACTTo classify the chemical form of stable carbon released from unirradiated stainless steel, which is the material used to simulate irradiated stainless steel, under highly alkaline and low-oxygen conditions, type 304 and 316 stainless-steel powders were immersed in 0.005 M NaOH solution. Gas and liquid samples were analyzed to identify the chemical form of carbon released from the stainless steel. The liquid samples were divided into unfiltered and filtered samples. In the gaseous phase, hydrocarbons such as methane and ethane were not detected. In the liquid phase, carboxylic acids (formic and acetic acids) were detected. However, the sum of the carbon concentrations of the carboxylic acids was significantly lower than the total organic carbon (TOC) concentration in the unfiltered samples. In the filtered samples, the TOC concentration was closer to the sum of the carbon concentrations than that for the unfiltered samples. In addition, the concentrations of the metallic elements (particularly Fe and Cr), which are the main constituents of the stainless steels, tended to decrease upon ultrafiltration. This suggests that the sorption of carbon on metallic compounds (e.g., colloidal iron hydroxide) may have occurred.

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19

GHAURI, I. M., NAVEED AFZAL, and N. A. ZYREK. "A STUDY OF STRESS RELAXATION RATE IN UN-IRRADIATED AND NEUTRON-IRRADIATED STAINLESS STEEL." Modern Physics Letters B 21, no. 05 (February 20, 2007): 295–301. http://dx.doi.org/10.1142/s021798490701258x.

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Stress relaxation rate in un-irradiated and neutron-irradiated 303 stainless steel was investigated at room temperature. The specimens were exposed to 100 mC, Ra-Be neutron source of continuous energy 2–12 MeV for a period ranging from 4 to 16 days. The tensile deformation of the specimens was carried out using a Universal Testing Machine at 300 K. During the deformation, straining was frequently interrupted by arresting the cross head to observe stress relaxation at fixed load. Stress relaxation rate, s, was found to be stress dependent i.e. it increased with increasing stress levels σ0 both in un-irradiated and irradiated specimens, however the rate was lower in irradiated specimens than those of un-irradiated ones. A further decrease in s was observed with increase in exposure time. The experiential decrease in the relaxation rate in irradiated specimens is ascribed to strong interaction of glide dislocations with radiation induced defects. The activation energy for the movement of dislocations was found to be higher in irradiated specimens as compared with the un-irradiated ones.

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20

Robertson, C., S. Poissonnet, and L. Boulanger. "Plasticity in ion-irradiated austenitic stainless steels." Journal of Materials Research 13, no. 8 (August 1998): 2123–31. http://dx.doi.org/10.1557/jmr.1998.0297.

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In an attempt to take advantage of charged particle irradiation for studying the effects of irradiation on the mechanical properties of metals, we developed an experimental procedure based on the combination of transmission electron microscopy (TEM) and submicron indentation of ion-implanted layers. We applied this technique to industrial 316L steel, irradiated with krypton ions up to 10 dpa at 873 K. A domain where the penetration range of the ions and the indentation depth are compatible has been identified. The indentation tests then yielded a good estimate of hardness and bulk modulus, while the TEM observations provide microstructural information in the plastic regime. It is shown that the combination of the two techniques is necessary for rationalizing the observed results.

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21

Russakova, Alyona, Darya Alontseva, and Tatyana Kolesnikova. "The Effect of Deformation and Irradiation with High-Energy Krypton Ions on the Structure and Phase Composition of Reactor Steels." Advanced Materials Research 702 (May 2013): 88–93. http://dx.doi.org/10.4028/www.scientific.net/amr.702.88.

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The paper presents some results of a complex research of 12Cr18Ni10Ti stainless steel in the initial, deformed and irradiated ( 8436Kr+14, E=130MeV, Fmax=9x1015 ions/сm2) states using magnetometry, X-ray diffraction (XRD) and scanning electron microscopy (SEM) with electron backscattered diffraction (EBSD – analysis). Application of the EBSD method revealed differences between the non-irradiated and irradiated 12Cr18Ni10Ti steel specimens consisting in the fact that in the surface layer of an irradiated sample α-and ε - phases are formed. It was established that the fluence value affects the amount of magnetic α-phase. The study of the martensite α-phase morphology showed that in the deformed steel specimens there is αʹ- martensite of two scale levels.

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22

KOYABU, Ken, Kyoichi ASANO, Hidenori TAKAHASHI, Hiroshi SAKAMOTO, Shohei KAWANO, Tomomi NAKAMURA, Tsuneyuki HASHIMOTO, et al. "Weldability of Neutron-Irradiated Stainless Steel and Nickel-Base Alloy." QUARTERLY JOURNAL OF THE JAPAN WELDING SOCIETY 18, no. 4 (2000): 606–16. http://dx.doi.org/10.2207/qjjws.18.606.

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23

Surh, Michael P., J. B. Sturgeon, and W. G. Wolfer. "Vacancy cluster evolution and swelling in irradiated 316 stainless steel." Journal of Nuclear Materials 328, no. 2-3 (July 2004): 107–14. http://dx.doi.org/10.1016/j.jnucmat.2004.03.005.

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24

Fujii, K., and K. Fukuya. "Irradiation-induced microchemical changes in highly irradiated 316 stainless steel." Journal of Nuclear Materials 469 (February 2016): 82–88. http://dx.doi.org/10.1016/j.jnucmat.2015.11.035.

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25

Hure, J., S. El Shawish, L. Cizelj, and B. Tanguy. "Intergranular stress distributions in polycrystalline aggregates of irradiated stainless steel." Journal of Nuclear Materials 476 (August 2016): 231–42. http://dx.doi.org/10.1016/j.jnucmat.2016.04.017.

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26

Reisman, D. B., W. G. Wolfer, A. Elsholz, and M. D. Furnish. "Isentropic compression of irradiated stainless steel on the Z accelerator." Journal of Applied Physics 93, no. 11 (June 2003): 8952–57. http://dx.doi.org/10.1063/1.1571969.

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27

Kohno, Yutaka, Akira Kohyama, and David S. Gelles. "Microstructural examination of FFTF-irradiated manganese-stabilized martensitic stainless steel." Journal of Nuclear Materials 179-181 (March 1991): 725–27. http://dx.doi.org/10.1016/0022-3115(91)90191-9.

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28

Du, Donghai, Kai Sun, and Gary S. Was. "IASCC of neutron irradiated 316 stainless steel to 125 dpa." Materials Characterization 173 (March 2021): 110897. http://dx.doi.org/10.1016/j.matchar.2021.110897.

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29

Phillips, Richard J., Michael J. Shane, and Jay A. Switzer. "Electrochemical and photoelectrochemical deposition of thallium(III) oxide thin films." Journal of Materials Research 4, no. 4 (August 1989): 923–29. http://dx.doi.org/10.1557/jmr.1989.0923.

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Thallium (III) oxide is a degenerate n-type semiconductor with high optical transparency and electrical conductivity. Films of thallium(III) oxide can be electrochemically deposited onto conducting and p-type semiconducting substrates, and photoelectrochemically deposited onto n-type semiconducting substrates. Films deposited at currents below the mass transport limit onto platinum or stainless steel were columnar, and the current efficiency on stainless steel was 103 ±2%. Dendritic films were deposited at mass-transport-limited currents. Films were deposited with thicknesses ranging from 0.1 μm on n-silicon, to 170 μm on stainless steel. The photoelectrochemically deposited films were “direct-written” onto n-silicon, since the material was deposited only at irradiated portions of the electrode. Thin films were grown by irradiating the n-silicon with 450 nm monochromatic light, since the light was strongly absorbed by the thallium(III) oxide. The most uniform thin films were deposited when the n-silicon was initially irradiated with a short pulse of high intensity light. The pulse apparently promoted instantaneous nucleation of a high density of thallium(III) oxide nuclei.

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30

Sartowska, Bożena, Marek Barlak, Lech Walis, and Wojciech Starosta. "Re-Melting Technique with High Intense Pulsed Plasma Beams Applied for Surface Modification of Steel - Own Investigations." Materials Science Forum 879 (November 2016): 1668–73. http://dx.doi.org/10.4028/www.scientific.net/msf.879.1668.

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The plasma beams were generated in a Rod Plasma Injector (RPI) operated in the Deposition by Pulsed Erosion (DPE) and Pulse Implantation Doping (DPE) modes. Samples of unalloyed and austenitic stainless steels were irradiated with short (μs scale) intense (energy density 2.0-5.0 J/cm2) pulses. The near surface layer - thickness in μm range - was melted and simultaneously doped with active element like nitrogen, cerium and lanthanum. Heating and cooling processes were of non-equilibrium type. The most important obtained results were: (i) austenitic structure was present in unalloyed steels after HIPPB modification processes and (ii) modified surface layers of austenitic stainless steel showed significant improvement of tribological properties and increase of high temperature oxidation resistance as compared with initial material.

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31

Kenik, E. A., J. Bentley, and N. D. Evans. "Application of electron energy loss spectroscopy to microanalysis of irradiated stainless steels." Proceedings, annual meeting, Electron Microscopy Society of America 49 (August 1991): 732–33. http://dx.doi.org/10.1017/s0424820100087975.

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Energy dispersive X-ray spectroscopy (EDXS) has limited application to microanalysis of radioactive materials because of degraded detector performance and the “intrinsic” spectrum associated with the radioactive decay. Electron energy loss spectroscopy (EELS) is not affected by specimen radioactivity and also offers the possibility of improved spatial resolution. Measurements of radiation-induced segregation (RIS) in irradiated stainless steels have been made by both techniques. Analytical electron microscopy was performed at 100 kV in a Philips EM400T/FEG, equipped with an EDAX 9100/70 EDXS system and a Gatan 666 parallel detection EELS (PEELS). Microanalysis was performed in the STEM mode (<2-nm-diam probe with >0.5 nA) with the same acquisition time (50 s) used for both techniques.Initial measurements were performed on an ion-irradiated modified type 316 stainless steel (designated LS1A), which had moderate-width (∼20 nm) RIS profiles at grain boundaries. Profiles measured by EDXS and PEELS match well and show chromium depletion and nickel enrichment (Fig. 1).

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32

Li, Xiao Wei, Chao Liang Xu, Ying Hui An, Xiang Bing Liu, Fei Xue, Yuan Fei Li, and Wang Jie Qian. "The Studies of Irradiation Hardening and RIS on IASCC of Reactor Internals Bolts." Key Engineering Materials 871 (January 2021): 92–97. http://dx.doi.org/10.4028/www.scientific.net/kem.871.92.

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IASCC of stainless steel has been the most important issue for internals BFBs. The inspection data analysis indicates that there is a closed relation between irradiation fluence and cracked BFBs distribution. Then the nanoindentation and 3DAP tests were carried out to study the hardening and radiation induced segregation (RIS) behaviors of the reactor internals stainless steel specimens irradiated with 6 MeV Xe ions at room temperature. It is indicated that higher irradiation damage will cause more significant hardening and RIS and consequently increase the IASCC susceptibility.

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33

Yao, Caizhen, Wei Gao, Yayun Ye, Yong Jiang, Shizhen Xu, and Xiaodong Yuan. "The formation of periodic micro/nano structured on stainless steel by femtosecond laser irradiation." International Journal of Modern Physics B 31, no. 16-19 (July 26, 2017): 1744004. http://dx.doi.org/10.1142/s0217979217440040.

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Stainless steel surface was irradiated by linear polarized laser (800 nm, 35 fs, 4 Hz and 0.7 J/cm2) with different pulse numbers. Environmental scanning electron microscope (ESEM/EDS) was used for detailed morphology, microstructure and composition studies. The wettability of irradiated steel surface was tested by Interface Tensiometer JC-2000X and compared with untreated stainless steel. Results showed that micro/nanostripes with different periods were formed. The period increased with the increasing pulse numbers from 450 nm for 90 pulses to 500 nm for 180 pulses. The orientation of those stripes was parallel with the laser beam polarization. Nanoparticles were observed on those periodic structures. EDS indicated that the atomic ratio of Cr increased and the atomic ratios of Fe and Ni decreased after laser irradiation, which may enhance the corrosion resistance due to the Cr-rich layer. The prepared structure exhibited hydrophobic property without further treatment. The formation mechanism of micro/nanoperiodic structures was also explored.

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34

Dai, Y., X. Jia, J. C. Chen, W. F. Sommer, M. Victoria, and G. S. Bauer. "Microstructure of both as-irradiated and deformed 304L stainless steel irradiated with 800 MeV protons." Journal of Nuclear Materials 296, no. 1-3 (July 2001): 174–82. http://dx.doi.org/10.1016/s0022-3115(01)00565-7.

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35

Jiao, Z., J. Hesterberg, and G. S. Was. "Effect of post-irradiation annealing on the irradiated microstructure of neutron-irradiated 304L stainless steel." Journal of Nuclear Materials 500 (March 2018): 220–34. http://dx.doi.org/10.1016/j.jnucmat.2017.12.030.

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36

Cookson, J. M., G. S. Was, and P. L. Andresen. "Oxide-Induced Initiation of Stress Corrosion Cracking in Irradiated Stainless Steel." CORROSION 54, no. 4 (April 1998): 299–312. http://dx.doi.org/10.5006/1.3284856.

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37

WEI XUE-QIN, ZHENG QI-GUANG, GU JIAN-HUI, and LI ZAI-GUANG. "ABNORMAL TEMPERATURE FLUCTUATION OF STAINLESS STEEL IRRADIATED BY CW CO2 LASER." Acta Physica Sinica 48, no. 12 (1999): 2246. http://dx.doi.org/10.7498/aps.48.2246.

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38

Chen, J., Y. Dai, F. Carsughi, W. F. Sommer, G. S. Bauer, and H. Ullmaier. "Mechanical properties of 304L stainless steel irradiated with 800 MeV protons." Journal of Nuclear Materials 275, no. 1 (October 1999): 115–18. http://dx.doi.org/10.1016/s0022-3115(99)00147-6.

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39

Teramoto, Tokuo, and Masakatsu Saito. "Fatigue strength for stainless steel irradiated by high power laser beam." Fusion Engineering and Design 9 (January 1989): 193–99. http://dx.doi.org/10.1016/s0920-3796(89)80033-x.

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40

Ioka, I., M. Futakawa, A. Naito, Y. Nanjyo, K. Kiuchi, and T. Naoe. "Mechanical characterization of austenitic stainless steel ion irradiated under external stress." Journal of Nuclear Materials 329-333 (August 2004): 1142–46. http://dx.doi.org/10.1016/j.jnucmat.2004.04.038.

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41

Maksimkin, O. P., K. V. Tsai, L. G. Turubarova, T. Doronina, and F. A. Garner. "Characterization of 08Cr16Ni11Mo3 stainless steel irradiated in the BN-350 reactor." Journal of Nuclear Materials 329-333 (August 2004): 625–29. http://dx.doi.org/10.1016/j.jnucmat.2004.04.102.

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42

Spencer, Rory P., and Eann A. Patterson. "Observations of fatigue crack behaviour in proton‐irradiated 304 stainless steel." Fatigue & Fracture of Engineering Materials & Structures 42, no. 9 (July 23, 2019): 2120–32. http://dx.doi.org/10.1111/ffe.13087.

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43

Vo, H. T., A. Reichardt, D. Frazer, N. Bailey, P. Chou, and P. Hosemann. "In situ micro-tensile testing on proton beam-irradiated stainless steel." Journal of Nuclear Materials 493 (September 2017): 336–42. http://dx.doi.org/10.1016/j.jnucmat.2017.06.026.

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44

Agamy, S. A., M. Y. Khalil, M. Y. Hamza, and W. A. Abou-Taleb. "Effects of Microstructural Defects on Ion-Irradiated Type-316 Stainless Steel." Isotopenpraxis Isotopes in Environmental and Health Studies 26, no. 6 (January 1990): 287–91. http://dx.doi.org/10.1080/10256019008624300.

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45

NOGAMI, Shuhei, Yuki SATO, and Akira HASEGAWA. "OS2504 Evaluation of Micro-crack Initiation Behavior in Irradiated Stainless Steel." Proceedings of the Materials and Mechanics Conference 2011 (2011): _OS2504–1_—_OS2504–2_. http://dx.doi.org/10.1299/jsmemm.2011._os2504-1_.

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46

Takagi, I., Y. Ueyama, T. Komura, M. Akiyoshi, T. Sasaki, K. Moritani, and H. Moriyama. "Hydrogen Trapping in Stainless Steel Irradiated by H and He Ions." Fusion Science and Technology 60, no. 4 (November 2011): 1523–26. http://dx.doi.org/10.13182/fst11-a12722.

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47

Chen, Jiachao, and Peter Jung. "Hardness of irradiated and helium implanted DIN 1.4914 martensitic stainless steel." Journal of Nuclear Materials 212-215 (September 1994): 559–63. http://dx.doi.org/10.1016/0022-3115(94)90122-8.

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48

Yamane, Toshimi, Keiichi Hirao, and Yoritoshi Minamino. "Stress corrosion cracking of neutron-irradiated type-316 stainless-steel weld." Journal of Materials Science Letters 5, no. 9 (September 1986): 943–45. http://dx.doi.org/10.1007/bf01729283.

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49

Kato, Takahiko, Kiyotomo Nakata, Jiro Kuniya, Soumei Ohnuki, and Heishichiro Takahashi. "Cavity formation by hydrogen injection in electron-irradiated austenitic stainless steel." Journal of Nuclear Materials 155-157 (July 1988): 856–60. http://dx.doi.org/10.1016/0022-3115(88)90429-1.

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50

Lee, G. G., H. H. Jin, K. Chang, B. H. Lee, and J. Kwon. "Atomistic Analysis Of Radiation-Induced Segregation In Ion-Irradiated Stainless Steel 316." Archives of Metallurgy and Materials 60, no. 2 (June 1, 2015): 1179–84. http://dx.doi.org/10.1515/amm-2015-0093.

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Abstract:

Abstract Stainless steel (SS) is a well-known material for the internal parts of nuclear power plants. It is known that these alloys exhibit radiation-induced segregation (RIS) at point defect sinks at moderate temperature, while in service. The RIS behavior of SS can be a potential problem by increasing the susceptibility to irradiation-assisted stress corrosion cracking. In this work, the RIS behavior of solute atoms at sinks in SS 316 irradiated with Fe4+ ions were characterized by atom probe tomography (APT). There were torus-shaped defects along with a depletion of Cr and enrichment of Ni and Si. These clusters are believed to be dislocation loops resulting from irradiation. The segregation of solutes was also observed for various defect shapes. These observations are consistent with other APT results from the literature. The composition of the clusters was analyzed quantitatively almost at the atomic scale. Despite the limitations of the experiments, the APT analysis was well suited for discovering the structure of irradiation defects and performing a quantitative analysis of RIS in irradiated specimens.

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