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Journal Articles

${it In situ}$ WB-STEM observation of dislocation loop behavior in reactor pressure vessel steel during post-irradiation annealing

Du, Y.*; Yoshida, Kenta*; Shimada, Yusuke*; Toyama, Takeshi*; Inoue, Koji*; Arakawa, Kazuto*; Suzudo, Tomoaki; Milan, K. J.*; Gerard, R.*; Onuki, Somei*; et al.

Materialia, 12, p.100778_1 - 100778_10, 2020/08

In order to ensure the integrity of the reactor pressure vessel in the long term, it is necessary to understand the effects of irradiation on the materials. In this study, irradiation-induced dislocation loops were observed in neutron-irradiated reactor pressure vessel specimens during annealing using our newly developed WB-STEM. It was confirmed that the proportion of $$<100>$$ loops increased with increasing annealing temperature. We also succeeded in observing the phenomenon that two $$frac{1}{2}$$$$<111>$$ loops collide into a $$<100>$$ loop. Moreover, a phenomenon in which dislocation loops decorate dislocations was also observed, and the mechanism was successfully explained by molecular dynamics simulation.

Journal Articles

Effects of one-dimensional migration of self-interstitial atom clusters on the decreasing behaviour of their number density in electron-irradiated $$alpha$$-iron

Abe, Yosuke; Sato, Yuki*; Hashimoto, Naoyuki*; Onuki, Somei*

Philosophical Magazine, 100(1), p.110 - 125, 2020/00

 Times Cited Count:2 Percentile:25.87(Materials Science, Multidisciplinary)

We derive analytical models associated with the experimentally revealed one-dimensional (1D) migration mechanisms to examine the decreasing behavior of the cluster number density. The model calculation indicates that the detrapping of the stationary SIA clusters causes the surface annihilation of the liberated SIA clusters, leading to the decrease in their number density. The decreasing behavior is in closer accordance with the experimental data when setting the impurity concentration in the same order as the estimation from the previous in situ HVEM experiment. This result suggests that the trapping and detrapping of the SIA clusters are the possible underlying processes for the decreasing behavior.

Journal Articles

Vacancy effects on one-dimensional migration of interstitial clusters in iron under electron irradiation at low temperatures

Sato, Yuki*; Abe, Yosuke; Abe, Hiroaki*; Matsukawa, Yoshitaka*; Kano, Sho*; Onuki, Somei*; Hashimoto, Naoyuki*

Philosophical Magazine, 96(21), p.2219 - 2242, 2016/06

 Times Cited Count:7 Percentile:45.78(Materials Science, Multidisciplinary)

We performed in situ observation of one-dimensional (1D) migration of self-interstitial atom (SIA) clusters in iron under electron irradiation at 110-300 K using high-voltage electron microscopy. Most 1D migration was stepwise positional changes of SIA clusters at irregular time intervals at all temperatures. The frequency of 1D migration did not depend on the irradiation temperature. It was directly proportional to the damage rate, suggesting that 1D migration was induced by electron irradiation. In contrast, the 1D migration distance depended on the temperature: distribution of the distance ranged over 100 nm above 250 K, decreased steeply between 250 and 150 K and was less than 20 nm below 150 K. The distance was independent of the damage rate at all temperatures. Next, we examined fluctuation in the interaction energy between an SIA cluster and vacancies of random distribution at concentrations $$10^{-4}$$-$$10^{-2}$$, using molecular statics simulations. The fluctuation was found to trap SIA clusters of 4 nm diameter at vacancy concentrations higher than $$10^{-3}$$. We proposed that 1D migration was interrupted by impurity atoms at temperatures higher than 250 K, and by vacancies accumulated at high concentration under electron irradiation at low temperatures where vacancies are not thermally mobile.

Journal Articles

Corrosion resistance of Al-alloying high Cr-ODS steels in stagnant lead-bismuth

Takaya, Shigeru; Furukawa, Tomohiro; Inoue, Masaki; Fujisawa, Toshiharu*; Okuda, Takanari*; Abe, Fujio*; Onuki, Somei*; Kimura, Akihiko*

Journal of Nuclear Materials, 398(1-3), p.132 - 138, 2010/03

 Times Cited Count:43 Percentile:94.69(Materials Science, Multidisciplinary)

Oxide dispersion strengthened (ODS) ferritic steels with excellent high-temperature strength are the candidates for fuel cladding tubes. But, the compatibility with lead bismuth eutectic (LBE) is one of the key issues in accelerator driven system and LBE cooled fast reactors. Addition of Al and increase in Cr may have beneficial influence on the compatibility. Addition of Al, however, causes a decrease in high-temperature strength. A significantly higher Cr concentration results in aging embrittlement. Therefore, we need to find their optimal amount to balance corrosion resistance with high-temperature strength. In this study, the cross sections of the samples after 3,000 h of exposure to LBE with 10$$^{-8}$$ wt% oxygen at 650 $$^{circ}$$C are examined in detail using scanning electron microscope and Auger electron spectroscopy. The observation shows that very thin Al oxide layer is formed continuously between multiple oxide layer/internal oxide zone and matrix, and that such Al oxide layer suppresses further growth of multiple oxide layer/internal oxide zone. The average oxide layer thickness shows a tendency to get thinner by increasing in Al content from about 2 to 4 wt%, although significant dependency on Cr content is not recognized. Furthermore, the additional corrosion test for 5,000 h is conducted. These materials show good corrosion resistance even after 5,000 h of exposure to LBE containing 10$$^{-6}$$ wt% at 650 $$^{circ}$$C. Addition of 3.5 wt% Al is very effective in improving corrosion resistance.

Journal Articles

High-temperature strength characterization of advanced 9Cr-ODS ferritic steels

Ukai, Shigeharu*; Otsuka, Satoshi; Kaito, Takeji; Sakasegawa, Hideo*; Chikada, Nobuyoshi*; Hayashi, Shigenari*; Onuki, Somei*

Materials Science & Engineering A, 510-511, p.115 - 120, 2009/06

 Times Cited Count:79 Percentile:95.88(Nanoscience & Nanotechnology)

Oxide dispersion strengthened (ODS) ferritic steels, which are the most promising candidate materials for advanced fast reactor fuel elements, have exceptional creep strength at 973 K. The superior creep property of 9Cr-ODS ferritic steels is ascribed to the formation of a nonequilibrium phase, designated as the residual ferrite. The yield strength of the residual ferrite itself has been determined to be as high as 1360 MPa at room temperature from nanoindentation measurements. The creep strength is enhanced by minimizing the number of packet boundaries induced by the martensitic phase transformation. The creep strain occurs by sliding at weaker regions such as at the grain boundaries and packet boundaries. It is found that 9Cr-ODS ferritic steels behave as fiber composite materials comprising the harder residual ferrite and the softer tempered martensite.

Journal Articles

Super ODS steels R&D for fuel cladding of next generation nuclear systems, 4; Mechanical properties at elevated temperatures

Furukawa, Tomohiro; Otsuka, Satoshi; Inoue, Masaki; Okuda, Takanari*; Abe, Fujio*; Onuki, Somei*; Fujisawa, Toshiharu*; Kimura, Akihiko*

Proceedings of 2009 International Congress on Advances in Nuclear Power Plants (ICAPP '09) (CD-ROM), p.9221_1 - 9221_7, 2009/05

As fuel cladding material for lead bismuth-cooled fast reactors and supercritical pressurized water-cooled fast reactors, our research group has been developing highly corrosion-resistant oxide dispersion strengthened ferritic steels with superior high-temperature strength. In this study, the mechanical properties of super ODS steel candidates at elevated temperature have been evaluated. Tensile tests, creep tests and low cycle fatigue tests were carried out for a total of 21 types of super ODS steel candidates which have a basic chemical composition of Fe-16Cr-4Al-0.1Ti-0.35Y$$_{2}$$O$$_{3}$$, with small variations. The testing temperatures were 700$$^{circ}$$C (for tensile, creep and low cycle fatigue tests) and 450$$^{circ}$$C (for tensile test). The major alloying parameters of the candidate materials were the compositions of Cr, Al, W and the minor elements such as Hf, Zr and Ce etc. The addition of the minor elements is considered effective in the control of the formation of the YAl complex oxides, which improves high-temperature strength. The addition of Al was very effective for the improvement of corrosion resistance. However, the addition also caused a reduction in high-temperature tensile strength. Among the efforts aimed at increasing high-temperature strength, such as the low-temperature hot-extrusion process, solution strengthening by W and the addition of minor elements, a remarkable improvement of strength was observed in ODS steel with a basic chemical composition of 2W-0.6Hf steel (SOC-14) or 2W-0.6Zr steel (SOC-16). The same behavior was also observed in creep tests, and the creep rupture times of SOC-14 and SOC-16 at 700$$^{circ}$$C - 100MPa were greater than 10,000 h. The strength was similar to that of no-Al ODS steels. No detrimental effect by the additional elements on low-cycle fatigue strength was observed in this study. These results showed that the addition of Hf/Zr to ODS-Al steels was effective in improving high-temperature strength.

Journal Articles

Super ODS steels R&D for fuel cladding of next generation nuclear systems, 1; Introduction and alloy design

Kimura, Akihiko*; Kasada, Ryuta*; Iwata, Noriyuki*; Kishimoto, Hirotatsu*; Zhang, C. H.*; Isselin, J.*; Dou, P.*; Lee, J. H.*; Muthukumar, N.*; Okuda, Takanari*; et al.

Proceedings of 2009 International Congress on Advances in Nuclear Power Plants (ICAPP '09) (CD-ROM), p.9220_1 - 9220_8, 2009/05

Cladding material development is essential for realization of highly efficient high burn-up operation of next generation nuclear systems, where high performance is required for the materials, that is, high strength at elevated temperature, high resistance to corrosion and high resistance to irradiation. Oxide dispersion strengthening (ODS) ferritic steels are considered to be most adequate for the cladding material because of their high strength at elevated temperature. In this work, "Super ODS steel" that has better corrosion resistance than 9Cr-ODS steel, has been developed for application to cladding of a variety of next generation nuclear systems. In the following ten papers, the recent experimental results of "Super ODS steel" R&D will be presented, indicating that many unexpected preferable features were found in the mechanical properties of nano-sized oxide dispersion high-Cr ODS ferritic steel. A series of paper begins with alloy design of "Super ODS steel". Corrosion issue requires Cr concentration more than 14wt.%, but aging embrittlement issue requires less than 16wt.%. An addition of 4wt.%Al is effective to improve corrosion resistance of 16wt.%Cr-ODS steel in supercritical water (SCW) and lead-bismuth eutectic (LBE), while it is detrimental to high-temperature strength. Additions of 2wt.%W and 0.1wt.%Ti are necessary to keep high strength at elevated temperatures. An addition of small amount of Zr or Hf results in a significant increase in creep strength at 700 $$^{circ}$$C in Al added ODS steels. Tube manufacturing was successfully done for the super ODS steel candidates. "Super ODS steel" is promising for the fuel cladding material of next generation nuclear systems, and the R&D is now ready to proceed to the next stage of empirical verification.

Journal Articles

Super ODS steels R&D for fuel cladding of next generation nuclear systems, 2; Effect of minor alloying elements

Onuki, Somei*; Hashimoto, Naoyuki*; Ukai, Shigeharu*; Kimura, Akihiko*; Inoue, Masaki; Kaito, Takeji; Fujisawa, Toshiharu*; Okuda, Takanari*; Abe, Fujio*

Proceedings of 2009 International Congress on Advances in Nuclear Power Plants (ICAPP '09) (CD-ROM), p.9306_1 - 9306_5, 2009/05

For development of advanced ferritic ODS steels including high concentration of Cr and Al, the effect of minor alloying elements on fine dispersion of oxide particle was investigated. Microstructural analysis for Fe-16Cr-4Al-mY$$_{2}$$O$$_{3}$$-nZr or mHf due to TEM indicated that 0.3Zr or 0.6Hf are the optimum concentration. The mechanism of nano-sized oxide formation was also discussed.

Journal Articles

Super ODS steels R&D for fuel cladding of next generation nuclear systems, 3; Development of high performance attrition type ball mill

Okuda, Takanari*; Fujiwara, Masayuki*; Nakai, Tatsuyoshi*; Shibata, Kenichi*; Kimura, Akihiko*; Inoue, Masaki; Ukai, Shigeharu*; Onuki, Somei*; Fujisawa, Toshiharu*; Abe, Fujio*

Proceedings of 2009 International Congress on Advances in Nuclear Power Plants (ICAPP '09) (CD-ROM), p.9229_1 - 9229_4, 2009/05

Oxygen content in ODS ferritic steel is the most important element to determine the mechanical properties. The oxygen contamination from the air is perfectly prevented by using new designed ball mill and the subsequent process control. Zr, Hf and Ti added ODS steels with three oxygen levels for the evaluation tests are fabricated.

Journal Articles

Super ODS steels R&D for fuel cladding of next generation nuclear systems, 6; Corrosion behavior in SCPW

Lee, J. H.*; Kimura, Akihiko*; Kasada, Ryuta*; Iwata, Noriyuki*; Kishimoto, Hirotatsu*; Zhang, C. H.*; Isselin, J.*; Dou, P.*; Muthukumar, N.*; Okuda, Takanari*; et al.

Proceedings of 2009 International Congress on Advances in Nuclear Power Plants (ICAPP '09) (CD-ROM), p.9223_1 - 9223_6, 2009/05

Corrosion is a critical issue for cladding materials, especially, in sever corrosion environment as supercritical pressurized water (SCPW). In this work, the effects of alloy elements on the corrosion behavior in SCPW were investigated for a series of oxide dispersion strengthened (ODS) steels to design alloy compositions for corrosion resistant super ODS ferritic steels. Corrosion tests were carried out for the ODS steels with different concentrations of Cr and Al in SCPW at 773 K at 25 MPa with 8 ppm of dissolved oxygen. The corrosion rate of SUS430, which contained 16wt.%Cr, was much higher than 16Cr-ODS steel. This suggests that nano-sized oxide particles dispersion and very fine grains play an important role in suppression of the corrosion. The corrosion of the ODS steel was reduced by an addition of Al in 16wt.%Cr-ODS steel but not in 19Cr-ODS steel. FE-EPMA chemical analysis clearly indicated that the surface of the Al added ODS steels was covered by alumina which suppresses the corrosion in SCPW. It is considered that an adequate combination of the contents of Cr and Al is ranging (14-16)Cr and (3.5-4.5)Al.

Journal Articles

Super ODS steels R&D for fuel cladding of next generation nuclear systems, 5; Mechanical properties and microstructure

Kasada, Ryuta*; Lee, S. G.*; Lee, J. H.*; Omura, Takamasa*; Zhang, C. H.*; Dou, P.*; Isselin, J.*; Kimura, Akihiko*; Inoue, Masaki; Ukai, Shigeharu*; et al.

Proceedings of 2009 International Congress on Advances in Nuclear Power Plants (ICAPP '09) (CD-ROM), p.9072_1 - 9072_5, 2009/05

The newly-developed Al-added ODS ferritic steels with an addition of Zr or Hf, socalled super ODS candidate steels, showed good notch-impact properties in the as-received condition with keeping the excellent creep strength.

Journal Articles

Super ODS steels R&D for fuel cladding of next generation nuclear systems, 7; Corrosion behavior and mechanism in LBE

Sano, Hiroyuki*; Fujisawa, Toshiharu*; Kimura, Akihiko*; Inoue, Masaki; Ukai, Shigeharu*; Onuki, Somei*; Okuda, Takanari*; Abe, Fujio*

Proceedings of 2009 International Congress on Advances in Nuclear Power Plants (ICAPP '09) (CD-ROM), p.9308_1 - 9308_5, 2009/05

Corrosion of structural materials is one of the serious problems when lead-bismuth eutectic alloy (LBE) is used as a coolant material in next generation nuclear systems. In this study, dissolution experiments of synthetic Fe-Cr-Al alloys and developed super ODS steel candidates into LBE under several partial pressures of oxygen were conducted. Dissolution behaviors of major components in such steels into LBE were investigated. Interfacial behavior between LBE and steels was also observed. In addition, partial potential diagrams of the Fe-Cr-Al-O system at several conditions were established as basic data. From the potential diagrams, the partial pressure range of oxygen was estimated for the stable protective oxide layer formation at the interface. At lower oxygen partial pressure than the pressure that is enough for the formation of the stable oxide layer, a rough oxide layer was formed at the interface in all samples, and the alloy elements dissolved into LBE through it. On the other hand, at the oxygen partial pressure to form stable oxide layer, a dense and very thin oxide layer was formed especially on the higher aluminum content steel, preventing the alloy dissolution into LBE. From the results, aluminum and chromium content in steel were very important for preventing the corrosion by LBE.

Journal Articles

Super ODS steels R&D for fuel cladding of next generation nuclear systems, 8; Ion irradiation effects at elevated temperatures

Kishimoto, Hirotatsu*; Kasada, Ryuta*; Kimura, Akihiko*; Inoue, Masaki; Okuda, Takanari*; Abe, Fujio*; Onuki, Somei*; Fujisawa, Toshiharu*

Proceedings of 2009 International Congress on Advances in Nuclear Power Plants (ICAPP '09) (CD-ROM), p.9219_1 - 9219_8, 2009/05

The Super ODS steels, having excellent high-temperature strength and highly corrosion resistant, are considered to increase the energy efficiency by higher temperature operation and extend the lifetime of next generation nuclear systems. High-temperature strength of the ODS steels strongly depends on the dispersion of oxide particles, therefore, the irradiation effect on the dispersed oxides is critical in the material development. In the present research, ion irradiation experiments were employed to investigate microstructural stability under the irradiation environment at elevated temperatures. Ion irradiation experiments were performed with 6.4 MeV Fe ions irradiated at 650 $$^{circ}$$C up to a nominal displacement damage of 60 dpa. Microstructural investigation was carried out using TEM and EDX. No significant change of grains and grain boundaries was observed by TEM investigation after the ion irradiation. Main oxide particles in the 16Cr-4Al-0.1Ti (SOC-1) ODS steel were (Y, Al) complex oxides. (Y, Ti) complex oxides were in 16Cr-0.1Ti (SOC-5) and 15.5Cr-2W-0.1Ti (SOCP-3). (Y, Zr) complex oxides were in 15.5Cr-4Al-0.6Zr (SOCP-1). No significant modification of these complex oxides was detected after the ion irradiation up to 60 dpa at 650 $$^{circ}$$C. The stable complex oxides are considered to keep highly microstructural stability of the Super ODS steels under the irradiation environments.

Journal Articles

Super ODS steels R&D for fuel cladding of next generation nuclear systems, 9; Damage structure evolution under electron-irradiation

Onuki, Somei*; Hashimoto, Naoyuki*; Ukai, Shigeharu*; Kimura, Akihiko*; Inoue, Masaki; Kaito, Takeji; Fujisawa, Toshiharu*; Okuda, Takanari*; Abe, Fujio*

Proceedings of 2009 International Congress on Advances in Nuclear Power Plants (ICAPP '09) (CD-ROM), p.9307_1 - 9307_4, 2009/05

The aim of this study is to survey microstructural properties and irradiation response of advanced high Cr and Al ODS steels. The effects of minor element addition and heat-treatment are also investigated. In these steels, black dots-like dislocation loops were formed around oxide particles during electron irradiation, and then the behavior depended on the type of additional elements, but the irradiation resistance was confirmed generally. The irradiation response was not sensitive as the heat-treatment, but the minor element addition (Zr and Hf) showed an intensive suppressing the loop growth. The results suggest that a large number of oxides enhanced the mutual recombination of the irradiation-induced point defects, especially at their surface.

Journal Articles

Super ODS steels R&D for fuel cladding of next generation nuclear systems, 10; Cladding tube manufacturing and summary

Ukai, Shigeharu*; Onuki, Somei*; Hayashi, Shigenari*; Kaito, Takeji; Inoue, Masaki; Kimura, Akihiko*; Fujisawa, Toshiharu*; Okuda, Takanari*; Abe, Fujio*

Proceedings of 2009 International Congress on Advances in Nuclear Power Plants (ICAPP '09) (CD-ROM), p.9232_1 - 9232_7, 2009/05

The super ODS cladding were manufactured into 8.5 mm outer diameter and 0.5 mm thickness by using pilger.mill rolling and intermediate heat treatment. The final heat treatment was successfully conducted at 1150 $$^{circ}$$C for 1h to make perfectly recrystallized structure. The manufactured cladding exhibits a secondary recrystallized grains with {110} $$<$$100$$>$$ Goss orientation. As a summary of super ODS steels R&D, the ODS steel, 16Cr-4Al-2W-0.1Ti added with small amount of Hf or Zr, is the prime candidate of "Super ODS steels" which has high strength at elevated temperatures, high resistance to corrosion for SCW and LBE, and high resistance to irradiation.

Journal Articles

Corrosion behavior of Al-alloying high Cr-ODS steels in lead-bismuth eutectic

Takaya, Shigeru; Furukawa, Tomohiro; Aoto, Kazumi; M$"u$ller, G.*; Weisenburger, A.*; Heinzel, A.*; Inoue, Masaki; Okuda, Takanari*; Abe, Fujio*; Onuki, Somei*; et al.

Journal of Nuclear Materials, 386-388, p.507 - 510, 2009/04

 Times Cited Count:44 Percentile:95.19(Materials Science, Multidisciplinary)

The corrosion resistance of ODS steels with 0$$sim$$3.5 wt% Al and 13.7$$sim$$17.3 wt% Cr and of a 12Cr steel were examined. The experiments were conducted at 550 and 650 $$^{circ}$$C up to 3,000 h in stagnant LBE containing 10$$^{-6}$$ and 10$$^{-8} $$wt% oxygen for the ODS steels and at 550 $$^{circ}$$C up to 5,000 h in stagnant LBE containing 10$$^{-8}$$ wt% oxygen for the 12Cr steel, respectively. Protective Al oxide scales were formed on the surfaces of ODS steels with about 3.5 wt% Al and 13.7$$sim$$17.3 wt% Cr. The addition of Al is very effective to improve the corrosion resistance of ODS steels. The ODS steel with 16 wt% Cr and no Al does not show any corrosion resistance except for the specimen exposed to LBE with 10$$^{-6}$$ wt% oxygen at 650 $$^{circ}$$C. It is not expected to improve the corrosion resistance by increasing solely Cr content.

Journal Articles

Extra-irradiation hardening of reduced activation ferritic/martensitic steel by multi-ion irradiation

Ando, Masami; Wakai, Eiichi; Okubo, Nariaki; Ogiwara, Hiroyuki; Sawai, Tomotsugu; Onuki, Somei*

Nihon Kinzoku Gakkai-Shi, 71(12), p.1107 - 1111, 2007/12

 Times Cited Count:1 Percentile:14.26(Metallurgy & Metallurgical Engineering)

no abstracts in English

Journal Articles

Structure of nano-size oxides in ODS steels and its stability under electron irradiation

Oka, Keiichiro*; Onuki, Somei*; Yamashita, Shinichiro; Akasaka, Naoaki; Otsuka, Satoshi; Tanigawa, Hiroyasu

Materials Transactions, 48(10), p.2563 - 2566, 2007/10

 Times Cited Count:29 Percentile:81.12(Materials Science, Multidisciplinary)

For understanding the microstructural details of nano-size oxide particles, three types of ODS ferritic and austenitic steels were examined by high voltage electron microscopy, EDS and AP-FIM. The oxide included Y, Ti and O and showed a shell-like structure with different composition. The shell-like structure depends on crystal structure of the matrix during fabrication process. To evaluate the irradiation stability of the oxide particles, the electron irradiation was carried out to 47 dpa in the temperature range between room temperature and 923 K. During the irradiation, the oxide particles did not show obvious change in size. The irradiation behavior is discussed comparing with the results recently reported.

Journal Articles

Microstructural development of a heavily neutron-irradiated ODS ferritic steel (MA957) at elevated temperature

Yamashita, Shinichiro; Akasaka, Naoaki; Ukai, Shigeharu; Onuki, Somei*

Journal of Nuclear Materials, 367-370(1), p.202 - 207, 2007/08

 Times Cited Count:64 Percentile:97.19(Materials Science, Multidisciplinary)

Microstructural observation was done on a neutron-irradiated oxide dispersion strengthened (ODS) ferritic steel, MA957. Since MA957 has been investigated from various viewpoints, special emphases in this study were laid on oxide behaviors including phase stability under irradiation at elevated temperature ($$sim$$973 K). Transmission electron microscopy (TEM) observation of the Y-Ti complex oxide particles showed they were fine ($$sim$$40 nm) whereas the Ti-oxide particles were relatively coarse ($$sim$$300 nm). Dispersion parameters of oxide particles, such as mean size and number density, changed due to irradiation. This fact implies that the recoil resolution of the oxide particles. When irradiated at 973 K, some Y-Ti complex oxides were surviving and interacted with the dislocation structures, which delayed the dislocation recovery and consequently stabilized the elongated grain structures. This is the first evidence showing that oxide particles are effectively functioning as pinning points of dislocations in motion under irradiation to a dose of $$sim$$100 dpa.

Journal Articles

Effect of helium and hydrogen production on irradiation hardening of F82H steel irradiated by ion beams

Wakai, Eiichi; Ando, Masami; Sawai, Tomotsugu; Onuki, Somei*

Materials Transactions, 48(6), p.1427 - 1430, 2007/06

 Times Cited Count:5 Percentile:40.6(Materials Science, Multidisciplinary)

Effects of helium and hydrogen production on irradiation hardening of martensitic steel F82H (Fe-8Cr-2W-0.2V-0.04Ta-0.1C) were examined by dual or triple beam experiments. The effects of tempering and cold working were also examined. The irradiations were performed at about 500$$^{circ}$$C to 50 dpa under simultaneous dual beams of 10.5 MeV-Fe and 1.05 MeV-He or triple beams of those and 380keV-H ions. The value of appm-He/dpa for the dual ion beams was about 15, and the values of appm-He/dpa and appm-H/dpa for the triple ion beams were 15 and 15 (or 150), respectively. The hardness of the irradiated specimens measured at room temperature using a micro indentation after the irradiations. Irradiation softening and hardening was observed in F82H-std, F82H+CW and a non-tempered F82H steels irradiated at about 500$$^{circ}$$C to 18 and 50 dpa, respectively, by dual ion beams. The hardness of the specimens irradiated at about 500$$^{circ}$$C to 18 dpa under triple ion beams was harder than that under dual ion beams.

62 (Records 1-20 displayed on this page)