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Journal Articles

Development of ODS tempered martensitic steel for high burn up fuel cladding tube of SFR

Otsuka, Satoshi; Tanno, Takashi; Oka, Hiroshi; Yano, Yasuhide; Tachi, Yoshiaki; Kaito, Takeji; Hashidate, Ryuta; Kato, Shoichi; Furukawa, Tomohiro; Ito, Chikara; et al.

2018 GIF Symposium Proceedings (Internet), p.305 - 314, 2020/05

Oxide Dispersion Strengthened (ODS) steel has been developed worldwide as a high-strength and radiation-tolerant steel used for advanced nuclear system. Japan Atomic Energy Agency (JAEA) has been developing ODS steel as the primary candidate material of Sodium cooled Fast Reactor (SFR) high burn-up fuel cladding tube. Application of high burn-up fuel to SFR core can contribute to improvement of economical performance of SFR in conjunction with volume and hazardousness reduction of radioactive waste. This paper described the current status and future prospects of ODS tempered martensitic steel development in JAEA for SFR fuel application.

Journal Articles

Hydrogen desorption spectra from excess vacancy-type defects enhanced by hydrogen in tempered martensitic steel showing quasi-cleavage fracture

Saito, Kei*; Hirade, Tetsuya; Takai, Kenichi*

Metallurgical and Materials Transactions A, 50(11), p.5091 - 5102, 2019/11

 Times Cited Count:1 Percentile:63.77(Materials Science, Multidisciplinary)

An attempt was made to separate and identify hydrogen peaks desorbed from lattice defects formed by plastic-strain in the presence of hydrogen in tempered martensitic steel showing quasi-cleavage fracture using thermal desorption spectroscopy from a low temperature (L-TDS) and positron annihilation spectroscopy (PAS). The L-TDS results made it possible to separate two peaks, namely, that of the original desorption and also that of new desorption. The PAS results revealed that the new desorption obtained by L-TDS corresponded to vacancy-type defects. Hydrogen enhanced vacancy-type defect concentration, approximately 10$$^{-5}$$ order in terms of atomic ratio, formed within 1.5 mm from the fracture surface, These results indicate that the accumulation of excess vacancy-type defects enhanced by hydrogen in the local region can lead to nanovoid nucleation and coalescence in plastic deformation, resulting in quasi-cleavage fracture of tempered martensitic steel.

Journal Articles

Study on simulation of thermal desorption spectra for a tempered martensitic steel with vacancies induced by hydrogen and strain

Ebihara, Kenichi; Saito, Kei*; Takai, Kenichi*

"Suiso Zeika No Kihon Yoin To Tokusei Hyoka" Kenkyukai Hokokusho, p.57 - 61, 2018/09

no abstracts in English

Journal Articles

Numerical simulation of hydrogen thermal desorption profile under assumption of two kinds of trap sites for tempered martensitic steel

Tsuchida, Yutaka*; Ebihara, Kenichi

Tetsu To Hagane, 103(11), p.653 - 659, 2017/11

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

A single peak in thermal desorption profiles of hydrogen, which are measured in low-temperature thermal desorption spectrometry (L-TDS) for a very thin plate specimen of tempered martensitic steel, was reproduced successfully by the superposition of two Gaussian distributions. Then, the parameters concerning the detrapping rate constants for both peaks, which are trap energy and pre-exponential factor, were calculated using the Choo-Lee plot. We confirmed that Kissinger model incorporating the obtained parameters could simulate the two peaks. In addition, we reproduced the single peak well using the reaction-diffusion equation incorporating the obtained parameters and the appropriate trap site concentration. From the results, we interpreted that the one peak corresponds to dislocation and the other to grain-boundary.

Journal Articles

Unusual plastic deformation behavior in lath martensitic steel containing high dislocation density

Harjo, S.; Kawasaki, Takuro; Tomota, Yo*; Gong, W.

Materials Science Forum, 905, p.46 - 51, 2017/08

Journal Articles

Reverse austenite transformation behavior in a tempered martensite low-alloy steel studied using ${it in situ}$ neutron diffraction

Tomota, Yo*; Gong, W.*; Harjo, S.; Shinozaki, Tomoya*

Scripta Materialia, 133, p.79 - 82, 2017/05

AA2017-0349.pdf:2.43MB

 Times Cited Count:13 Percentile:25.04(Nanoscience & Nanotechnology)

Journal Articles

Dislocation characteristics in lath martensitic steel by neutron diffraction

Harjo, S.; Kawasaki, Takuro; Gong, W.; Aizawa, Kazuya

Journal of Physics; Conference Series, 746(1), p.012046_1 - 012046_7, 2016/10

BB2015-1438.pdf:0.82MB

 Times Cited Count:3 Percentile:13.22

Journal Articles

Study on modeling of thermal desorption spectra of hydrogen including variation of vacancy-type trap sites

Ebihara, Kenichi; Saito, Kei*; Takai, Kenichi*

"Suiso Zeika No Kihon Yoin To Tokusei Hyoka Kenkyukai Chukan Hokokukai" Shimposium Yokoshu (USB Flash Drive), p.30 - 35, 2016/09

no abstracts in English

Journal Articles

Presence of $$varepsilon$$-martensite as an intermediate phase during the strain-induced transformation of SUS304 stainless steel

Hatano, Masaharu*; Kubota, Yoshiki*; Shobu, Takahisa; Mori, Shigeo*

Philosophical Magazine Letters, 96(6), p.220 - 227, 2016/06

 Times Cited Count:6 Percentile:55.22(Materials Science, Multidisciplinary)

We have investigated the formation process of $$alpha$$'-martensite from the $$gamma$$-phase induced by external strain using in-situ synchrotron diffraction experiments, combined with Lorentz transmission electron microscopy (TEM) and high-resolution TEM observations. It is clearly demonstrated that $$varepsilon$$-martensite with hexagonal symmetry appears as an intermediate structure during the plastic deformation of SUS304 stainless steel. In addition to stacking faults and dislocations, interfaces between the twin structures presumably play a key role in the formation of $$varepsilon$$-martensite.

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:38 Percentile:6.89(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

Fatigue properties of F82H irradiated at 523 K to 3.8 dpa

Miwa, Yukio; Jitsukawa, Shiro; Yonekawa, Minoru

Journal of Nuclear Materials, 329-333(Part2), p.1098 - 1102, 2004/08

 Times Cited Count:11 Percentile:37.43(Materials Science, Multidisciplinary)

Fatigue properties were examined on a reduced activation ferritic/martensitic steel, and preliminary results were presented. F82H steel was irradiated at 523 K to 3.8 dpa, and then fatigue-tested at 298-573 K in vacuum with total strain range of 0.4-1.0%. Effect of irradiation on fatigue lives was observed on test at 298 K with total strain range of 0.4%. The fatigue life of irradiated specimen was reduced to about 1/7 of unirradiated specimen. The reduction of the fatigue life was attributed to the occurrence of channel fracture. Effect of test temperature was discussed.

Journal Articles

Material issues of blanket systems for fusion reactors; Compatibility with cooling water

Miwa, Yukio; Tsukada, Takashi; Jitsukawa, Shiro

Purazuma, Kaku Yugo Gakkai-Shi, 80(7), p.551 - 557, 2004/07

Environmental assisted cracking (EAC) is one of the materials issues for the reactor core components of light water power reactors (LWRs). Much experience and knowledge have been obtained about EAC in LWR field. They will be useful to manage the EAC of water-cooled blanket systems of the fusion reactors. For the austenitic stainless steels and the reduced-activation ferritic/martensitic steels, they clarifies that the EAC of a water-cooled blanket does not seem to be critical issues. However some uncertainties about influences of water temperatures, water chemistries and stress conditions may affect on the EAC. Considerations and further investigations investigating for such the uncertainties are discussed.

Journal Articles

Microstructural study of irradiated isotopically tailored F82H steel

Wakai, Eiichi; Miwa, Yukio; Hashimoto, Naoyuki*; Robertson, J. P.*; Klueh, R. L.*; Shiba, Kiyoyuki; Abiko, Kenji*; Furuno, Shigemi*; Jitsukawa, Shiro

Journal of Nuclear Materials, 307-311(Part.1), p.203 - 211, 2002/12

 Times Cited Count:20 Percentile:21.12

no abstracts in English

Journal Articles

Deformation behavior of high temperature shape memory alloys before and after neutron irradiation

Hoshiya, Taiji

Kagaku To Kogyo, 75(8), p.379 - 387, 2001/08

no abstracts in English

Journal Articles

Radiation effects on the plasticity and microstructure of Ti-Al-V alloys containing $$beta$$ phase

Sawai, Tomotsugu; Tabuchi, Masayuki*; Wakai, Eiichi; Hishinuma, Akimichi

Proceedings of Materials Research Society Symposium, Vol.650, p.R3.9.1 - R3.9.6, 2001/00

no abstracts in English

Journal Articles

Microstructure of welded and thermal-aged low activation steel F82H IEA heat

Sawai, Tomotsugu; Shiba, Kiyoyuki; Hishinuma, Akimichi

Journal of Nuclear Materials, 283-287(Part.1), p.657 - 661, 2000/12

 Times Cited Count:35 Percentile:9.93

no abstracts in English

Journal Articles

Tensile results of low-activation martensitic steel irradiated in HFIR RB-11J and RB-12J spectrally tailored capsules

Shiba, Kiyoyuki; Klueh, R. L.*; Miwa, Yukio; Igawa, Naoki; Robertson, J. P.*

Fusion Materials Semiannual Progress Report (DOE/ER-0313/28), p.131 - 135, 2000/06

no abstracts in English

JAEA Reports

The evaluation of material base standard of ODS ferritic stainless steel core component for fast breeder reactors

Mizuta, Shunji; ;

JNC-TN9400 2000-048, 28 Pages, 2000/04

JNC-TN9400-2000-048.pdf:0.64MB

ODS (Oxide Dispersion Strengthened) ferritic-martainsitic steels are one of the most prospective cladding materials for advanced fast breeder reactors, since they are expected to have excellent swelling resistance and superior high temperature strength due to the finely distributed stable oxide particles(Y$$_{2}$$O$$_{3}$$). Properties and the tentative strength equations for ODS ferritic-martainsitic were proposed on the basis of the latest data to apply to the feasibility study of the sodium coolant MOX fuel plant. The items of equations are follows. (1)creep rupture strength (2)correction factor of creep rupture strength (in Na and in reactor) (3)outer surface eorrosion (Na) (4)inner surface corrosion (in MOX fuel pin) (5)thermal conductivity

JAEA Reports

Evaluation of charpy impact property in high strength ferritic/martensitic steel (PNC-FMS)

;

JNC-TN9400 2000-035, 164 Pages, 2000/03

JNC-TN9400-2000-035.pdf:3.67MB

High Strength Ferritic/Martensitic Steel (PNC-FMS : 0.12C-11Cr-0,5Mo-2W-0.2V-0.05Nb), developed by JNC, is one of the candidate materials for the long-life core of large-scale fast breeder reactor. Ductile brittle transition temperature (DBTT) was tentatively determined in 1992 in material design base standard of PNC-FMS. Howevcr, specimen size effect on impact property and upper shelf energy (USE) have not been evaluated. ln this study, effects of specimen size, thermal aging and neutron irradiation on the charpy impact property of PNC-FMS were evaluated, using together with recently obtained data. The design value of USE and DBTT as fabricated and each correlation of aging and irradiation effects were determined. The results are summarized as follows. (1)lt was found that USE is related to (Bb) as USE=m(Bb)$$^{n}$$, where B is specimen width, b is ligament size and both m and n are constant. For PNC-FMS, n value is equal to 1.4. It's possible to determine n value from USE (J) for full size specimen using the correlation: n=1.38$$times$$10$$^{-3}$$ USE + 1.20. (2)lt was clarified that DBTT is correlated with (BKt) as DBTT=p(log$$_{10}$$BKt)+q, where Kt is elastic stress concentration factor and both p and q are constant. For PNC-FMS, the correlation is as follows: DBTT=119(log$$_{10}$$BKt)-160. (3)DBTT estimated at the irradiation temperature from 350 to 650 $$^{circ}$$C for sub size specimen (width and height are 3 and 10 mm, respectively), was below 180 $$^{circ}$$C, based on the design value of DBTT as fabricated and each correlation of aging and irradiation effects.

Journal Articles

Current status and future R&D for reduced-activation ferritic/martensitic steels

Hishinuma, Akimichi; Koyama, Akira*; R.L.Klueh*; D.S.Gelles*; Ehrlich, K.*; W.Dietz*

Journal of Nuclear Materials, 258-263, p.193 - 204, 1998/00

 Times Cited Count:183 Percentile:0.17

no abstracts in English

34 (Records 1-20 displayed on this page)