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

https://doi.org/10.1016/j.jnucmat.2013.05.063

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

https://doi.org/10.1016/j.fusengdes.2011.06.005

Aging properties of reduced activation ferritic/martensitic steel F82H was researched at temperature ranging from 400C to 650C up to 100,000 hr. Microstructure, tensile, and Charpy properties were carried out. Laves was found at temperatures between 550 and 650C and MC carbides were found at the temperatures between 500 and 600C over 10,000 hr. These precipitates caused degradation in toughness, especially at temperatures ranging from 550C to 650C. Tensile properties do not have serious aging effect, except for 650C, 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 650C 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 550C.

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

https://doi.org/10.1557/opl.2011.394

The irradiation behaviour in two different precipitation hardening types of Ni-base alloys with the ultra high purity grade (EHP), namely, the ' 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 or Ni) at 673 K. Triple ion-beam irradiation tests were conducted up to 90 dpa (by Ni, 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(') 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 planes. Those dominating defect structures at 823 K by triple ion-beam are classified as followings, bimodal distributions in EHP('), 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('), 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.

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 AlO 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.

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

https://doi.org/10.1016/j.jnucmat.2007.03.212

A reduced activation ferritic/martensitic steel, F82H has been tested through slow strain rate tests with strain rates of 310 s 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 C to 550 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.

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

https://doi.org/10.1016/j.jnucmat.2007.04.021

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, 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 300C 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.

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

http://www-naweb.iaea.org/napc/physics/FEC/FEC2006/html/node165.htm#34754

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.

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

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

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

https://doi.org/10.2320/jinstmet.71.244

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 AlO-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.

Radiation hardening and -embrittlement due to He production in F82H steel irradiated at 250 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

https://doi.org/10.1016/j.jnucmat.2004.10.167

The dependence of helium production on radiation-hardening and -embrittlement has been examined in a reduced-activation martensitic F82H steel doped with B, B and B+B irradiated at 250C 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 irradiation was also performed to implant about 85 appm He atoms at 120C 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 DBTT for the F82H+B, F82H+B+B, and F82H+B were 40, 110, and 155C, respectively. The shifts of DBTT due to He production were evaluated as about 70C by 150 appmHe and 115C by 300 appmHe. The DBTT shift in the small punch testing was evaluated as 50C.

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

https://doi.org/10.1016/j.jnucmat.2004.08.001

F82H (Fe-8Cr-2W) and its variant doped with 2%Ni were irradiated up to 20 dpa at 300C 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 300C. The strain hardening of Ni doped specimens was, however, strongly affected by irradiation. Results provide basics to determine allowable stress levels at temperatures below 400C.