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Journal Articles

Water corrosion resistance of ODS ferritic-martensitic steel tubes

Narita, Takeshi; Ukai, Shigeharu; Kaito, Takeji; Otsuka, Satoshi; Matsuda, Yasushi*

Journal of Nuclear Science and Technology, 45(2), p.99 - 102, 2008/02

 Times Cited Count:5 Percentile:34.43(Nuclear Science & Technology)

The oxide dispersion strengthened (ODS) ferritic-martensitic steels are being developing for application as advanced fast reactor cladding and fusion blanket materials, in order to allow increased operation temperature. Water corrosion test of ODS ferritic-martensitic steels was conducted under a controlled alkali water environment to evaluate the water corrosion behavior, comparing to conventional 17 mass% Cr austenitic stainless steel (PNC316) and 11 mass% Cr ferritic-martensitic stainless steel (PNC-FMS). It was showed that 9Cr-ODS martensitic steel and 12Cr-ODS ferritic steel have superior water corrosion resistance, and comparable to that of PNC316 and PNC-FMS at 333K for 1,000h under varying pH of 8.4, 10, 12.

Journal Articles

Irradiation effects on precipitation and its impact on the mechanical properties of reduced-activation ferritic/martensitic steels

Tanigawa, Hiroyasu; Sakasegawa, Hideo; Hashimoto, Naoyuki*; Klueh, R. L.*; Ando, Masami; Sokolov, M. A.*

Journal of Nuclear Materials, 367-370(1), p.42 - 47, 2007/08

 Times Cited Count:26 Percentile:83.63(Materials Science, Multidisciplinary)

It was previously reported that reduced-activation ferritic/martensitic steels (RAFs), such as F82H-IEA and its heat treatment variant, ORNL9Cr-2WVTa, JLF-1 and 2%Ni-doped F82H, showed a variety of changes in ductile-brittle transition temperature (DBTT) and yield stress after irradiation at 573K up to 5dpa. These differences could not be interpreted solely as an effect of irradiation hardening caused by dislocation loop formation. To address these observations, the precipitation behavior of the irradiated steels was examined by weight analysis, X-ray diffraction analysis and chemical analysis on extraction residues. The results suggested that irradiation affects precipitation as if it was forced to reach the thermal equilibrium state at irradiation temperature 573K, which usually never be achieved by aging. The details of precipitates in the irradiated RAFs were examined to determine their impact on the mechanical properties, which obtained by tensile, Charpy impact, and bend bar toughness tests. Transmission electron microscopy was performed on thin films and extraction replica specimens to analyze the size distribution, chemical composition and crystal structure of precipitates. It turned out that the hardening level normalized by square root of average packet size showed a linear dependence on the increase of extracted precipitate weight. This dependence suggests that the difference in irradiation hardening between RAFs was caused by the different precipitation behavior on packet, block and prior austenitic grain boundaries during irradiation. The simple Hall-Petch law could be applicable to interpret this dependence. Detailed analytical results will be presented and their interpretation discussed.

Journal Articles

Radiation induced phase instability of precipitates in reduced-activation ferritic/martensitic steels

Tanigawa, Hiroyasu; Sakasegawa, Hideo; Ogiwara, Hiroyuki*; Kishimoto, Hirotatsu*; Koyama, Akira*

Journal of Nuclear Materials, 367-370(1), p.132 - 136, 2007/08

 Times Cited Count:46 Percentile:93.05(Materials Science, Multidisciplinary)

It was previously reported that reduced-activation ferritic/martensitic steels (RAFs), such as F82H-IEA and JLF-1, showed a variety of changes in its mechanical property after neutron irradiation at 573K up to 5dpa, and have possible correlation with precipitation. The effects of irradiation on precipitation were also reported previously. In this study, irradiation effects on precipitation were investigated in detail utilizing ion irradiation in which irradiation condition could be controlled with high accuracy. F82H IEA heat, JLF-1 HFIR heat, and aged F82H-IEA (873K$$times$$30k h) were used for experiments. The specimens were irradiated at DuET facility, Inst. of Advanced Energy, Kyoto University up to 10 dpa at 573K with 6.4MeV Fe$$^{3+}$$ ion. Cross sectional TEM thin film specimens of ion irradiated region were made utilizing focused ion beam (FIB) processor with micro-sampling system at JAERI. These thin film specimens were made to contain both irradiated region and non-irradiated region beneath irradiated region. Size distribution and aspect ratio of precipitates were analyzed on each region. It turned out that the finer precipitates were dominant in irradiated region of F82H compared to that in non-irradiated region, but fewer and larger precipitates were dominant in irradiated region of JLF-1. These results confirmed the presence of irradiation effects on precipitate evolution even at 573K, which was observed in neutron irradiated RAFs.

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:18.41(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:19 Percentile:76.17(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

Properties of friction welds between 9Cr-ODS martensitic and ferritic-martensitic steels

Uwaba, Tomoyuki; Ukai, Shigeharu; Nakai, Tatsuyoshi*; Fujiwara, Masayuki*

Journal of Nuclear Materials, 367-370(2), p.1213 - 1217, 2007/08

 Times Cited Count:16 Percentile:71.35(Materials Science, Multidisciplinary)

Friction welding between oxide-dispersion-strengthened-steel (ODS) and ferritic-martensitic 11Cr-0.5Mo-2W, V, Nb steel (PNC-FMS) was examined to investigate the feasibility of the welding of this alloy combination. Forge pressure was varied in the welding test to investigate the effect on the soundness of the weld joint. Post weld heat treatment was necessary for the weld joint because the heat affected zone caused quench hardening in the as-welded condition. The welds were cross-sectioned and examined to determine their metallurgical structure, hardness and tensile strength. Since the heat treatment produced a softening zone near the weld interface, the furnace- cooling rate was investigated at which the softening did not occur. The weld joints were cold rolled and the soundness of the weld interface was determined by the tensile test with the cold rolled weld joints.

Journal Articles

Nano-mesoscopic structural characterization of 9Cr-ODS martensitic steel for improving creep strength

Otsuka, Satoshi; Ukai, Shigeharu; Sakasegawa, Hideo; Fujiwara, Masayuki; Kaito, Takeji; Narita, Takeshi

Journal of Nuclear Materials, 367-370(1), p.160 - 165, 2007/08

 Times Cited Count:60 Percentile:95.79(Materials Science, Multidisciplinary)

This paper describes the effect on creep strength and microstructure of 9Cr-oxide dispersion strengthened martensitic steel (9Cr-ODS steel) brought by the differences in titanium concentration and consolidation temperature. The increase of titanium concentration to 0.30-0.35wt% was shown to provide remarkable improvement of creep strength accompanied by the increase of residual-alpha ferrite. The elevation of hot-extrusion temperature notably degraded the creep strength, however, appeared to increase the volume fraction of residual-alpha ferrite. Creep deformation process of 9Cr-ODS steel was discussed to explain these results based on microstructure observations.

Journal Articles

Mechanisms of retention and blistering in near-surface region of tungsten exposed to high flux deuterium plasmas of tens of eV

Shu, Wataru; Luo, G.; Yamanishi, Toshihiko

Journal of Nuclear Materials, 367-370(2), p.1463 - 1467, 2007/08

 Times Cited Count:69 Percentile:96.78(Materials Science, Multidisciplinary)

The mechanisms of retention and blistering in the near-surface region of tungsten exposed to high glux deuterium plasmas of tens of eV were studied with a variety of techniques, such as XRD, TEM, SEM, TDS, NRA and ERDA. The results of small angle XRD at a fixed incident angle of 1.5$$^{circ}$$ indicated that within the experimental error there was a zero change in the lattice parameter after the plasma exposure. This implies that deuterium does not exist in the lattice interstitial sites, but instead forms a deuterium-vacancy complex and then clusters and further bubbles (deuterium molecules in vacancy clusters and voids) in near-surface region. Cross-sectional TEM observations showed that small blisters with a diameter of around 30 nm and nano-cracks formed in the near-surface region before the formation of larger blisters with diameters of up to a few microns (comparable to grain size). The TDS results strongly indicated that deuterium existed in the molecular form in tungsten after the plasma exposure. The NRA and ERDA results suggest that the maximum atomic ratio of deuterium against tungsten reached as high as 1-2% in the near-surface region. These evidences suggest that crystal defects like vacancies should be generated due to lowering of the formation energy of vacancies by the intrusion of a great number of hydrogen isotope atoms into the near-surface region of tungsten.

Journal Articles

Recent results on beryllium and beryllides in Japan

Mishima, Yoshinao*; Yoshida, Naoaki*; Kawamura, Hiroshi; Ishida, Kiyohito*; Hatano, Yuji*; Shibayama, Tamaki*; Munakata, Kenzo*; Sato, Yoshiyuki*; Uchida, Munenori*; Tsuchiya, Kunihiko; et al.

Journal of Nuclear Materials, 367-370(2), p.1382 - 1386, 2007/08

 Times Cited Count:28 Percentile:85.23(Materials Science, Multidisciplinary)

no abstracts in English

Journal Articles

Effect of temperature change on the irradiation hardening of the structural alloys for ITER blanket and ITER TBM irradiated to 1.5 dpa in JMTR

Jitsukawa, Shiro; Wakai, Eiichi; Okubo, Nariaki; Omi, Masao

Journal of Nuclear Materials, 367-370(1), p.539 - 543, 2007/08

 Times Cited Count:2 Percentile:18.41(Materials Science, Multidisciplinary)

By using an irradiation capsule with temperature control capability independent from reactor power, a reduced activation ferritic steel and austenitic stainless steels were irradiated at temperatures of 250 and 350$$^{circ}$$C. Irradiation was performed to the accumulated damage level of 1.5 dpa. For some of the specimens, temperature was changed during irradiation. The temperature change reduced the irradiation hardening of the austenitic steel. On the other hand, it caused slightly increased hardening for a reduced activation ferritic steel. The mechanism of the observed temperature change effect, and the impact of the extra-hardening on the residual ductility is discussed.

Journal Articles

Status of reduced activation ferritic/martensitic steel development

Baluc, N.*; Gelles, D. S.*; Kimura, Akihiko*; Jitsukawa, Shiro; Klueh, R. L.*; Odette, G. R.*; Van der Schaaf, B.*; Jinnan, Y.*

Journal of Nuclear Materials, 367-370(1), p.33 - 41, 2007/08

 Times Cited Count:213 Percentile:99.78(Materials Science, Multidisciplinary)

Recent research results obtained in Europe, Japan, China and the USA on reduced-activation ferritic/martensitic (RAFM) steels are reviewed. The present status of different RAFM steel products (plate, powder HIPped steel, many types of fusion and diffusion welds, unirradiated and irradiated states) is sufficient to present a strong case for the use of the steels in ITER test blanket modules. For application in DEMO, more research is needed, including the use of the International Fusion Materials Irradiation Facility (IFMIF) in order to quantify the effects of large amounts of transmutation products, such as helium and hydrogen.

Journal Articles

Luminescence of Cr-doped alumina induced by charged particle irradiation

Inoue, Aichi; Nagata, Shinji*; To, Kentaro*; Tsuchiya, Bun*; Yamamoto, Shunya; Shikama, Tatsuo*

Journal of Nuclear Materials, 367-370(2), p.1112 - 1116, 2007/08

 Times Cited Count:5 Percentile:36.73(Materials Science, Multidisciplinary)

The ion-induced luminescence behavior has been investigated for Cr-doped alumina (ruby, $$alpha$$-Al$$_{2}$$O$$_{3}$$: 0.5 wt% Cr) under MeV region H and He ion irradiation. The initial yield of the R-line luminescence (693 nm) increased proportionally to the projected range of the incident H ions, indicating that the ion- induced luminescence efficiency for an H ion was constant with independence of the electronic energy loss. However, the He induced luminescence yield was not directly related to either the projected range or the incident energy, corresponding that the efficiency decreased with increasing the electronic energy loss. The reduction of R-line yields by the ion irradiation was adequately explained by decreasing R-line luminescence centers.

Journal Articles

Development of 300$$^{circ}$$C heat resistant boron-loaded resin for neutron shielding

Morioka, Atsuhiko; Sakurai, Shinji; Okuno, Koichi*; Sato, Satoshi; Verzirov, Y. M.; Kaminaga, Atsushi; Nishitani, Takeo; Tamai, Hiroshi; Shibama, Yusuke; Yoshida, Shigeru*; et al.

Journal of Nuclear Materials, 367-370(2), p.1085 - 1089, 2007/08

 Times Cited Count:25 Percentile:83.02(Materials Science, Multidisciplinary)

A 300$$^{circ}$$C heat-resistant neutron shielding material is newly developed, which consists of phenol-based resin with 6 weight-% boron. The neutron shielding performance of the developed resin, examined by the $$^{252}$$Cf neutron source, is almost the same as that of the polyethylene. The neutron shielding characteristic was also estimated by 3D Monte Carlo Code MCNP-4C2 using the continuous energy cross section data sets based on the JENDL-3.2. The calculation result agrees well with the experimental result. To understand the kinds of the outgas from the developed resin in the high temperature region, the mass spectrum of the outgas was measured until $$sim$$300$$^{circ}$$C by Thermal Desorption Specroscopy (TDS). The observed mass number was 2, 17, 18, 28, 32, and 44. The number corresponds to hydrogen (H$$_{2}$$), ammonia (NH$$_{3}$$), water (H$$_{2}$$O), carbon monoxide (CO), oxygen (O$$_{2}$$), carbon dioxide (CO$$_{2}$$), respectively. The main outgas component from the resin at 100$$sim$$150$$^{circ}$$C was NH $$_{3}$$ and H$$_{2}$$O. The outgas of NH$$_{3}$$ and H$$_{2}$$O from the resin have been measured, however, the neutron shielding performance of the resin after 200$$^{circ}$$C baking was almost the same as that before baking. The quantitative analysis of the outgas from the resin in the high temperature region was done by the Temperature Programmed Desorption (TPD) / Gas Chromatography and Mass spectrometry (GC/MS). The 13 kinds of organic gases have been observed by the amount of $$mu$$g/g at 300$$^{circ}$$C. The neutron shielding performance of the developed resin at 300$$^{circ}$$C was simulated by the 3D analysis. The resonance cross section of the nucleus is broad at the high temperature region by the Doppler effect. The calculation results using 327$$^{circ}$$C library and 20$$^{circ}$$C library are almost same.

Journal Articles

Dynamics of deuterium implanted in boron coating film for wall conditioning

Nakahata, Toshihiko*; Yoshikawa, Akira*; Oyaizu, Makoto*; Oya, Yasuhisa*; Ishimoto, Yuki*; Kizu, Kaname; Yagyu, Junichi; Ashikawa, Naoko*; Nishimura, Kiyohiko*; Miya, Naoyuki; et al.

Journal of Nuclear Materials, 367-370(2), p.1170 - 1174, 2007/08

 Times Cited Count:3 Percentile:25.12(Materials Science, Multidisciplinary)

Retention and desorption behavior of deuterium implanted into pure boron films has been studied by means of the secondary ion mass spectroscopy. It was found that the factor dominating deuterium desorption was the sample temperature. At stage 1, below 573 K, the desorption of deuterium from B-D-B bond dominated and diffusion was the rate-determining process in this stage. Above 573 K, deuterium was mainly desorbed from B-D bonds, and recombination was the rate-determining process in this stage. The effective molecular recombination rate constant of deuterium trapped as B-D bond was determined by an isothermal annealing experiment.

Journal Articles

Effect of solute elements in Ni alloys on blistering under He$$^{+}$$ D$$^{+}$$ ion irradiation

Wakai, Eiichi; Ezawa, Tadashi*; Takenaka, Tsuyoshi*; Imamura, Junko*; Tanabe, Tetsuo*; Oshima, Ryuichiro*

Journal of Nuclear Materials, 367-370(1), p.478 - 482, 2007/08

 Times Cited Count:7 Percentile:46.54(Materials Science, Multidisciplinary)

no abstracts in English

Journal Articles

Effect of heat treatments on tensile properties of F82H steel irradiated by neutrons

Wakai, Eiichi; Ando, Masami; Sawai, Tomotsugu; Tanigawa, Hiroyasu; Taguchi, Tomitsugu; Stroller, R. E.*; Yamamoto, Toshio; Kato, Yoshiaki; Takada, Fumiki

Journal of Nuclear Materials, 367-370(1), p.74 - 80, 2007/08

 Times Cited Count:10 Percentile:57.51(Materials Science, Multidisciplinary)

no abstracts in English

Journal Articles

Mechanical properties of SiC/SiC composite with magnesium-silicon oxide interphase

Igawa, Naoki; Taguchi, Tomitsugu; Yamada, Reiji; Ishii, Yoshinobu; Jitsukawa, Shiro

Journal of Nuclear Materials, 367-370(1), p.725 - 729, 2007/08

 Times Cited Count:16 Percentile:67.69(Materials Science, Multidisciplinary)

no abstracts in English

Journal Articles

The Feasibility of recycling and clearance of active materials from fusion power plants

Zucchetti, M.*; El-Guebaly, L. A.*; Forrest, R. A.*; Marshall, T. D.*; Taylor, N. P.*; Tobita, Kenji

Journal of Nuclear Materials, 367-370(2), p.1355 - 1360, 2007/08

 Times Cited Count:20 Percentile:77.48(Materials Science, Multidisciplinary)

no abstracts in English

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:22 Percentile:80.08(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

Tensile and transient burst properties of advanced ferritic/martensitic steel claddings after neutron irradiation

Yano, Yasuhide; Yoshitake, Tsunemitsu; Yamashita, Shinichiro; Akasaka, Naoaki; Onose, Shoji; Takahashi, Heishichiro*

Journal of Nuclear Materials, 367-370(1), p.127 - 131, 2007/08

 Times Cited Count:11 Percentile:60.39(Materials Science, Multidisciplinary)

The effects of fast neutron irradiation on tensile and transient burst properties of advanced ferritic/martensitic steel claddings were investigated. Specimens were irradiated in the experimental fast reactor JOYO using the material irradiation rig at temperatures between 773 and 1013 K to fast neutron doses ranging from 11 to 102 dpa. The post-irradiation tensile and temperature-transient-to-burst tests were carried out. The results of mechanical tests showed that there was no significant degradation in tensile and transient burst strengths after neutron irradiation below 873 K. This was attributed to grain boundary strengthening caused by precipitates that preferentially formed on prior-austenite grain boundaries. Both strengths at neutron irradiation above about 903 K up to 102 dpa decreased due to recovery of lath martensite structures and recrystallization.

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