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Journal Articles

Irradiation response in weldment and HIP joint of reduced activation ferritic/martensitic steel, F82H

Hirose, Takanori; Sokolov, M. A.*; Ando, Masami; Tanigawa, Hiroyasu; Shiba, Kiyoyuki; Stoller, R. E.*; Odette, G. R.*

Journal of Nuclear Materials, 442(1-3), p.S557 - S561, 2013/11

 Times Cited Count:5 Percentile:51.14(Materials Science, Multidisciplinary)

Journal Articles

Microstructural evolution and void swelling in extra high purity Ni-base superalloy under multi-ion irradiation

Kim, G.; Shiba, Kiyoyuki; Sawai, Tomotsugu; Ioka, Ikuo; Kiuchi, Kiyoshi; Nakayama, Jumpei*

Proceedings of 2nd International Workshop on Structural Materials for Innovative Nuclear Systems (SMINS-2), p.273 - 279, 2012/12

Journal Articles

Long-term properties of reduced activation ferritic/martensitic steels for fusion reactor blanket system

Shiba, Kiyoyuki; Tanigawa, Hiroyasu; Hirose, Takanori; Sakasegawa, Hideo; Jitsukawa, Shiro

Fusion Engineering and Design, 86(12), p.2895 - 2899, 2011/12

 Times Cited Count:26 Percentile:7.59(Nuclear Science & Technology)

Aging properties of reduced activation ferritic/martensitic steel F82H was researched at temperature ranging from 400$$^{circ}$$C to 650$$^{circ}$$C up to 100,000 hr. Microstructure, tensile, and Charpy properties were carried out. Laves was found at temperatures between 550 and 650$$^{circ}$$C and M$$_{6}$$C carbides were found at the temperatures between 500 and 600$$^{circ}$$C over 10,000 hr. These precipitates caused degradation in toughness, especially at temperatures ranging from 550$$^{circ}$$C to 650$$^{circ}$$C. Tensile properties do not have serious aging effect, except for 650$$^{circ}$$C, which caused large softening even after 10.000 hr. Increase of precipitates also causes some degradation in ductility, but it is not critical. Large increase in DBTT caused by the large Laves phase precipitation at grain boundary was observed in the 650$$^{circ}$$C aging. Laves precipitates at grain boundary also degrades the USE of the aged materials. These aging test results provide F82H can be used up to 30,000 hr at 550$$^{circ}$$C.

Journal Articles

Status and key issues of reduced activation ferritic/martensitic steels as the structural material for a DEMO blanket

Tanigawa, Hiroyasu; Shiba, Kiyoyuki; M$"o$slang, A.*; Stoller, R. E.*; Lindau, R.*; Sokolov, M. A.*; Odette, G. R.*; Kurtz, R. J.*; Jitsukawa, Shiro

Journal of Nuclear Materials, 417(1-3), p.9 - 15, 2011/10

 Times Cited Count:99 Percentile:0.38(Materials Science, Multidisciplinary)

ITER construction was started, and R&D toward DEMO shifted to more practical stage. On this stage, the candidate material for DEMO blanket have to be the one which have sound engineering bases to be ready for engineering designing activity for DEMO reactor in 10 years. Reduced activation ferritic/martensitic (RAFM) steels, such as F82H (Fe-8Cr-2W-0.2V-0.04Ta) or EUROFER97 (Fe-9Cr-1W-0.2V- 0.12Ta), is the only material which currently have enough potential to meet this requirement, and selected as the target material in the R&D on materials engineering for DEMO blanket under the International Fusion Energy Research Centre (IFERC) project in the Broader Approach (BA) activities between EU and Japan. In this paper, current status of RAFM R&D is overviewed especially on fabrication technology, inspection/testing technology, and material database. Overview on irradiation effect study is also provided.

Journal Articles

Irradiation behavior of precipitation hardened Ni-base super-alloys with EHP grade under multi-ion irradiation

Kim, G.; Shiba, Kiyoyuki; Sawai, Tomotsugu; Ioka, Ikuo; Kiuchi, Kiyoshi

Materials Research Society Symposium Proceedings, Vol.1298, p.61 - 66, 2011/04

The irradiation behaviour in two different precipitation hardening types of Ni-base alloys with the ultra high purity grade (EHP), namely, the $$gamma$$' type and G phase type was investigated by multi-ion beam techniques simulated to the irradiation conditions in fuel cladding tubes used in sodium cooled FBRs. Single ion-beam irradiation tests were conducted up to 90 dpa (by Fe$$^{3+}$$ or Ni$$^{3+}$$) at 673 K. Triple ion-beam irradiation tests were conducted up to 90 dpa (by Ni$$^{3+}$$, 90 appmHe and 1350 appmH) at 823 K. The irradiation behaviour was examined by nano-indentation tests to irradiation hardening, and the microscopic observation by TEM to the distribution of dislocations, cavities and voids. The behaviour was compared with those of PNC316. The dominating irradiation defects in EHP($$gamma$$') alloy at 673 K by single ion-beam are Frank loops, perfect unfaulted loops and line dislocations. Whereas, those of EHP(WSi) alloy are the irradiation-induced G phase precipitates along ${111}$ planes. Those dominating defect structures at 823 K by triple ion-beam are classified as followings, bimodal distributions in EHP($$gamma$$'), bubbles in EHP(WSi) and voids in PNC316. The ratio of void swelling is estimated as nearly 0.01% in EHP(WSi), 0.2% in EHP($$gamma$$'), 3.4% in PNC316. From those results, the excellent irradiation properties of EHP(WSi) alloy is clarified as the inhibition effects of secondary irradiation defects.

Journal Articles

Irradiation effects on reduced activation ferritic/martensitic steels; Tensile, impact, fatigue properties and modelling

Jitsukawa, Shiro; Suzuki, Kazuhiko; Okubo, Nariaki; Ando, Masami; Shiba, Kiyoyuki

Nuclear Fusion, 49(11), p.115006_1 - 115006_8, 2009/11

 Times Cited Count:9 Percentile:60.56(Physics, Fluids & Plasmas)

Irradiation often causes hardening and reduction of elongation as well as toughness degradation to a considerable degree. Data, however, indicate that these changes remain in manageable ranges for ITER-TBM application. Moreover, the saturation tendency of the changes with neutron dose suggests that some of the reduced activation martensitic steels are feasible even for future DEMO applications. It is also stressed that the development of a design methodology that is compatible with the large irradiation induced changes is essential to enable these applications. Modeling activities for the macroscopic mechanical response are expected to play key roles in design methodology development. Macroscopic models of plasticity (a constitutive equation) and cyclic softening behavior after irradiation are discussed. Significance of models to estimate microstructural changes during irradiation and beneficial effects of the heat treatment for irradiation performance are also introduced.

Journal Articles

Effect of Ta rich inclusions and microstructure change during precracking on bimodal fracture of reduced-activation ferritic/martensitic steels observed in transition range

Tanigawa, Hiroyasu; Sokolov, M. A.*; Sawahata, Atsushi*; Hashimoto, Naoyuki*; Ando, Masami; Shiba, Kiyoyuki; Enomoto, Masato*; Klueh, R. L.*

Journal of ASTM International (Internet), 6(5), 10 Pages, 2009/05

The master curve (MC) method works when the transition fracture toughness values follow the MC, and once the value is scaled properly, the MC is usually independent of the type of steel or the type of test specimen. This method is very much depending on the assumption that the fracture initiation points are homogeneously distributed and its initiation mechanism is independent on test temperature. The reduced-activation ferritic/martensitic steels (RAFs), such as F82H (Fe-8Cr-2W-0.2V-0.04Ta), has Al$$_{2}$$O$$_{3}$$ Ta(V,Ti)O composite inclusions, or simple Ta(V)O inclusions, and shows inhomogeneous distribution, and it was revealed that that RAFs which contain Ta could initiate the facture in the different mechanism at lower temperature as the composite inclusions become fragile, and this should be considered when the toughness measured with small size toughness specimen which is usually tested at lower temperature.

Journal Articles

Technical issues of reduced activation ferritic/martensitic steels for fabrication of ITER test blanket modules

Tanigawa, Hiroyasu; Hirose, Takanori; Shiba, Kiyoyuki; Kasada, Ryuta*; Wakai, Eiichi; Serizawa, Hisashi*; Kawahito, Yosuke*; Jitsukawa, Shiro; Kimura, Akihiko*; Kono, Yutaka*; et al.

Fusion Engineering and Design, 83(10-12), p.1471 - 1476, 2008/12

 Times Cited Count:73 Percentile:1.48(Nuclear Science & Technology)

Reduced activation ferritic/martensitic steels (RAFMs) are recognized as the primary candidate structural materials for fusion blanket systems. F82H, which were developed and studied in Japan, was designed with an emphasis on high temperature properties and weldability. The database on F82H properties is currently the most extensive available among the existing RAFMs. The objective of this paper is to review the R&D status of F82H and to identify the key technical issues for the fabrication of an ITER Test Blanket Module (TBM) suggested by recent achievements in Japan.

Journal Articles

Stress corrosion cracking susceptibility of ferritic/martensitic steel in super critical pressurized water

Hirose, Takanori; Shiba, Kiyoyuki; Enoeda, Mikio; Akiba, Masato

Journal of Nuclear Materials, 367-370(2), p.1185 - 1189, 2007/08

 Times Cited Count:18 Percentile:19.94(Materials Science, Multidisciplinary)

A reduced activation ferritic/martensitic steel, F82H has been tested through slow strain rate tests with strain rates of 3$$times$$10$$^{-7}$$ s$$^{-1}$$ in Super Critical Pressurized Water (SCPW) environment. The water was pressurized up to 23.5 MPa and the range of its temperature was from 280 $$^{circ}$$C to 550 $$^{circ}$$C. The stress drop and the loss of ductility, both due to the stress corrosion cracking, have not been observed in all the specimens. Also the fracture surface showed no brittle fracture. The weight change during the SSRT depends strongly on the test temperature, but dissolved oxygen content does not have significant effects. The time dependence of weight change has been described through the plot of some parabolic curves.

Journal Articles

Thermo-structural analysis and design consideration of the replaceable backwall in IFMIF liquid lithium target

Nakamura, Hiroo; Ida, Mizuho; Chida, Teruo; Shiba, Kiyoyuki; Shimizu, Katsusuke*; Sugimoto, Masayoshi

Journal of Nuclear Materials, 367-370(2), p.1543 - 1548, 2007/08

 Times Cited Count:2 Percentile:79.17(Materials Science, Multidisciplinary)

The IFMIF is an accelerator-based intense neutron source for testing candidate materials of fusion reactor. Intense neutrons are emitted inside the Li flow through a backwall. The backwall made of 316L stainless steel or RAFM is attached to the target assembly with a lip seal welded by YAG laser. Since the backwall is operating under a severe neutron irradiation of 50 dpa/year and a maximum nuclear heating rate of 25 W/cm$$^{3}$$, thermo-structural design is one of critical issues in a target design. Thermal stress was calculated by ABAQUS code. As a permissible stress, yield strength at 300$$^{circ}$$C was used. In a case of the 316 stainless steel backwall, a maximum thermal stress was more than the permissible stress(164MPa). On the other hand, in case of the F82H backwall, a maximum thermal stress is was below the permissible stress(455MPa). Therefore, F82H is recommended as a backwall material.

Journal Articles

Creep behavior of reduced activation ferritic/martensitic steels irradiated at 573 and 773K up to 5dpa

Ando, Masami; Li, M.*; Tanigawa, Hiroyasu; Grossbeck, M. L.*; Kim, S.-W.; Sawai, Tomotsugu; Shiba, Kiyoyuki; Kono, Yutaka*; Koyama, Akira*

Journal of Nuclear Materials, 367-370(1), p.122 - 126, 2007/08

 Times Cited Count:13 Percentile:28.77(Materials Science, Multidisciplinary)

Irradiation creep behavior of the F82H and several JLF-1 steels have been measured up to 5 dpa, using helium-pressurized creep tubes irradiated in HFIR. These tubes were pressurized with helium to hoop stress levels of 0 to 400 MPa for irradiation temperature. The results of F82H and JLF-1 with 200 MPa hoop stress showed small creep strains ($$<$$ 0.15%) after irradiation. Irradiation creep rate in these steels is linearly dependent on the applied stress less than 200 MPa. However, at higher hoop stress level, the creep rate of them is nonlinear. The creep compliance coefficient for F82H and JLF-1 at 300 $$^{circ}$$C is very small values. These data contribute to a part of materials database for ITER Test blanket design work.

Journal Articles

Irradiation creep behavior of reduced activation ferritic/martensitic steel irradiated in HFIR

Ando, Masami; Tanigawa, Hiroyasu; Shiba, Kiyoyuki; Jitsukawa, Shiro; Kono, Yutaka*; Koyama, Akira*; Li, M.*; Stoller, R. E.*

Nippon Kinzoku Gakkai-Shi, 71(7), p.559 - 562, 2007/07

 Times Cited Count:0 Percentile:100(Metallurgy & Metallurgical Engineering)

no abstracts in English

Journal Articles

Status and key issues of reduced activation martensitic steels as the structural materials of ITER test blanket module and beyond

Tanigawa, Hiroyasu; Shiba, Kiyoyuki; Hirose, Takanori; Kasada, Ryuta*; Wakai, Eiichi; Jitsukawa, Shiro; Kimura, Akihiko*; Koyama, Akira*

Proceedings of 21st IAEA Fusion Energy Conference (FEC 2006) (CD-ROM), 6 Pages, 2007/03

The status of research and development of reduced activation martensitic steels (RAMs) in Japan are reviewed and key issues suggested from recent achievements in Japan since the last conference are highlighted, with the aim of the fabrication for the ITER Test Blanket Module (TBM) and application for the DEMO reactor. It was pointed out that international collaboration would be desirable for research on key issues such as precipitate stability under irradiation or Ta effects which are common for all RAMs and require an extensive research effort.

Journal Articles

Influence of Ti on inclusion formation of reduced activation ferritic/martensitic steels

Sawahata, Atsushi; Tanigawa, Hiroyasu; Shiba, Kiyoyuki; Enomoto, Masato*

Nippon Kinzoku Gakkai-Shi, 71(2), p.244 - 248, 2007/02

 Times Cited Count:1 Percentile:85.39(Metallurgy & Metallurgical Engineering)

Reduced activation ferritic/martensitic steels (RAFs), such as F82H (Fe-8Cr-2W-0.2V-0.04Ta-0.1C), are one of the leading candidates for structural materials of fusion reactors, and it is essential for RAFs R&D to assure its good toughness property for fusion application. In this study, the influence of Ti on impact property was studied based on microstructural analyses and charpy impact tests which were performed against general-purity F82H (0.004Ti-0.0060N) and high-purity F82H ($$<$$0.001Ti- 0.0014N). In general-purity F82H, its impact property around DBTT showed both 100% brittle fracture and brittle-ductile, and this tendency is not appeared in high-purity F82H. SEM observation on those brittle fracture surfaces of general-purity F82H revealed the presence of Al$$_{2}$$O$$_{3}$$-Ta(V,Ti)O complex oxides at the fracture initiation point. The size distribution analyses of oxides suggest that the complex oxide in general-purity F82H showed a higher number density than in high-purity F82H. In addition to this, EDS analyses showed the complex oxides in general-purity F82H had a strong peak of Ti, but they were not detected in the oxide in high-purity F82H. These results suggest the influence of Ti on complex oxide formation which affects impact property.

Journal Articles

Evaluation of creep properties of reduced activation ferritic steels

Nakata, Toshiya; Tanigawa, Hiroyasu; Shiba, Kiyoyuki; Komazaki, Shinichi*; Fujiwara, Mikio*; Kono, Yutaka*; Koyama, Akira*

Nippon Kinzoku Gakkai-Shi, 71(2), p.239 - 243, 2007/02

 Times Cited Count:3 Percentile:68.5(Metallurgy & Metallurgical Engineering)

no abstracts in English

Journal Articles

Evaluation of high-temperature tensile properties of reduced activation ferritic steels by small punch test

Nakata, Toshiya; Komazaki, Shinichi*; Nakajima, Motoki*; Kono, Yutaka*; Tanigawa, Hiroyasu; Shiba, Kiyoyuki; Koyama, Akira*

Nippon Kinzoku Gakkai-Shi, 70(8), p.642 - 645, 2006/08

 Times Cited Count:4 Percentile:62.72(Metallurgy & Metallurgical Engineering)

no abstracts in English

Journal Articles

Radiation hardening and -embrittlement due to He production in F82H steel irradiated at 250 $$^{circ}$$C in JMTR

Wakai, Eiichi; Jitsukawa, Shiro; Tomita, Hideki*; Furuya, Kazuyuki; Sato, Michitaka*; Oka, Keiichiro*; Tanaka, Teruyuki*; Takada, Fumiki; Yamamoto, Toshio*; Kato, Yoshiaki; et al.

Journal of Nuclear Materials, 343(1-3), p.285 - 296, 2005/08

 Times Cited Count:37 Percentile:7.03(Materials Science, Multidisciplinary)

The dependence of helium production on radiation-hardening and -embrittlement has been examined in a reduced-activation martensitic F82H steel doped with $$^{10}$$B, $$^{11}$$B and $$^{10}$$B+$$^{11}$$B irradiated at 250$$^{circ}$$C to 2.2 dpa. The total amounts of doping boron were about 60 massppm. The range of He concentration produced in the specimens was from about 5 to about 300 appm. Tensile and fracture toughness tests were performed after neutron irradiation. 50 MeV-He$$^{2+}$$ irradiation was also performed to implant about 85 appm He atoms at 120$$^{circ}$$C by AVF cyclotron to 0.03 dpa, and small punch testing was performed to obtain DBTT. Radiation-hardening of the neutron-irradiated specimens increased slightly with increasing He production. The 100 MPam$$^{1/2}$$ DBTT for the F82H+$$^{11}$$B, F82H+$$^{10}$$B+$$^{11}$$B, and F82H+$$^{10}$$B were 40, 110, and 155$$^{circ}$$C, respectively. The shifts of DBTT due to He production were evaluated as about 70$$^{circ}$$C by 150 appmHe and 115$$^{circ}$$C by 300 appmHe. The DBTT shift in the small punch testing was evaluated as 50$$^{circ}$$C.

Journal Articles

Effect of heat treatments on mechanical properties and microstructures of 8Cr-2W(F82H) steel doped with boron or boron and nitrogen

Wakai, Eiichi; Sato, Michitaka*; Okubo, Nariaki; Sawai, Tomotsugu; Shiba, Kiyoyuki; Jitsukawa, Shiro

Nippon Kinzoku Gakkai-Shi, 69(6), p.460 - 464, 2005/06

 Times Cited Count:1 Percentile:82.49(Metallurgy & Metallurgical Engineering)

no abstracts in English

Journal Articles

Post irradiation plastic properties of F82H derived from the instrumented tensile tests

Taguchi, Tomitsugu; Jitsukawa, Shiro; Sato, Michitaka*; Matsukawa, Shingo*; Wakai, Eiichi; Shiba, Kiyoyuki

Journal of Nuclear Materials, 335(3), p.457 - 461, 2004/12

 Times Cited Count:10 Percentile:40.16(Materials Science, Multidisciplinary)

F82H (Fe-8Cr-2W) and its variant doped with 2%Ni were irradiated up to 20 dpa at 300$$^{circ}$$C in the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory. Post irradiation tensile testing was performed at room temperature. During testing, the images of the specimens including the necked region were continuously recorded. Tests on cold worked material were also carried out for comparison. From the load-displacement curves and the strain distributions obtained from the images, flow stress levels and strain hardening behavior was evaluated. A preliminary constitutive equation for the plastic deformation of irradiated F82H is presented. The results suggest that the irradiation mainly causes defect-induced hardening while it did not strongly affect strain hardening at the same flow stress level for F82H irradiated at 300$$^{circ}$$C. The strain hardening of Ni doped specimens was, however, strongly affected by irradiation. Results provide basics to determine allowable stress levels at temperatures below 400$$^{circ}$$C.

Journal Articles

Effects of heat treatment process for blanket fabrication on mechanical properties of F82H

Hirose, Takanori; Shiba, Kiyoyuki; Sawai, Tomotsugu; Jitsukawa, Shiro; Akiba, Masato

Journal of Nuclear Materials, 329-333(Part1), p.324 - 327, 2004/08

 Times Cited Count:51 Percentile:4.52(Materials Science, Multidisciplinary)

Reduced activation ferritic/martensitic steel, RAFs is the leading candidates for the structural materials of breeding blankets. HIP is examined as a near-net-shape fabrication process for this structure. The HIP requires heating above the normalizing temperature and the final microstructural features depends on the HIP processing conditions. Conventional HIP process caused a prior-austenite grain (PAG) coarsening of RAFs and subsequent increase of ductile brittle transition temperature. Japanese RAFs F82H and its modified steels were investigated by metallurgical method after isochronal heat treatment up to 1473K simulating HIP equivalent thermal hysteresis. Although Conventional F82H IEA heat showed significant grain growth after conventional solid HIP conditions (1313K $$times$$ 2hr.), F82H with 0.1wt.% tantalum kept fine grain after the same heat treatment. On the other hands, conventional RAF/Ms with coarse grain were recovered by the post HIP normalizing at temperature below TaC dissolution temperature. This process can refine the PAG size of F82H more than ASTM grain size number 7.

58 (Records 1-20 displayed on this page)