Narita, Takeshi; Ukai, Shigeharu; Kaito, Takeji; Otsuka, Satoshi; Matsuda, Yasushi*
Journal of Nuclear Science and Technology, 45(2), p.99 - 102, 2008/02
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.
Nakazawa, Tetsuya; Naito, Akira*; Aruga, Takeo; Grismanovs, V.*; Chimi, Yasuhiro; Iwase, Akihiro*; Jitsukawa, Shiro
Journal of Nuclear Materials, 367-370(2), p.1398 - 1403, 2007/08
no abstracts in English
Tsuchiya, Kunihiko; Kawamura, Hiroshi; Ishida, Takuya
Journal of Nuclear Materials, 367-370(2), p.1018 - 1022, 2007/08
no abstracts in English
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
no abstracts in English
Shu, Wataru; Luo, G.; Yamanishi, Toshihiko
Journal of Nuclear Materials, 367-370(2), p.1463 - 1467, 2007/08
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 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.
Hirose, Takanori; Shiba, Kiyoyuki; Enoeda, Mikio; Akiba, Masato
Journal of Nuclear Materials, 367-370(2), p.1185 - 1189, 2007/08
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.
Yano, Yasuhide; Yoshitake, Tsunemitsu; Yamashita, Shinichiro; Akasaka, Naoaki; Onose, Shoji; Takahashi, Heishichiro*
Journal of Nuclear Materials, 367-370(1), p.127 - 131, 2007/08
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.
Kubota, Naoyoshi; Kondo, Keitaro; Ochiai, Kentaro; Nishitani, Takeo
Journal of Nuclear Materials, 367-370(2), p.1596 - 1600, 2007/08
The Neutron Elastic Recoil Detection Analysis (NERDA) using a neutron beam was proposed to extend the analyzing depth of hydrogen isotopes up to several hundreds micrometers. The 14.1 MeV neutrons beam produced by the Fusion Neutronics Source facility in Japan Atomic Energy Agency entered a sample from the normal direction. Emitted particles from the sample were measured using a E-E counter telescope detector. Incident neutron fluence was monitored with a U fission chamber located behind the target chamber. The proof-of-principle experiment was performed using a standard sample of deuterated polyethylene film containing a known concentration of deuterium with the thickness of 100 m. The depth resolution was evaluated to be 99 m corresponding to 12% of the maximum probing depth of 801 m for the sample. For a carbon-based Plasma Facing Component (PFC) sample the depth resolution was expected to be 61 m, which was enough to reveal hydrogen isotope distributions of co-deposited layers. Also, we applied NERDA to analyses of hydrogen isotope distributions in PFCs of JT-60U.
Kinjo, Tomohiro*; Nishikawa, Masabumi*; Enoeda, Mikio
Journal of Nuclear Materials, 367-370(2), p.1361 - 1365, 2007/08
no abstracts in English
Okubo, Nariaki; Wakai, Eiichi; Matsukawa, Shingo*; Sawai, Tomotsugu; Kitazawa, Sin-iti; Jitsukawa, Shiro
Journal of Nuclear Materials, 367-370(1), p.107 - 111, 2007/08
no abstracts in English
Okubo, Nariaki; Wakai, Eiichi; Tomita, Takeshi; Jitsukawa, Shiro
Journal of Nuclear Materials, 367-370(1), p.112 - 116, 2007/08
no abstracts in English
Yamashita, Shinichiro; Akasaka, Naoaki; Ukai, Shigeharu; Onuki, Somei*
Journal of Nuclear Materials, 367-370(1), p.202 - 207, 2007/08
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 (973 K). Transmission electron microscopy (TEM) observation of the Y-Ti complex oxide particles showed they were fine (40 nm) whereas the Ti-oxide particles were relatively coarse (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 100 dpa.
Ida, Mizuho; Nakamura, Hiroo; Sugimoto, Masayoshi
Journal of Nuclear Materials, 367-370(2), p.1557 - 1561, 2007/08
In the International Fusion Materials Irradiation Facility (IFMIF), radioactive nuclides are generated through the deuteron-lithium reaction and the neutron irradiation on the target vessel made of stainless steel. Beryllium-7 is the most dominant nuclide affecting accessibility to and maintenance scenario of IFMIF lithium loop.Dose equivalent rate around typical component of the lithium loop was calculated employing a code QAD-CGGP2R that can deal with three-dimensional (3-D) model of activated objects and radiation shield. Deposition rate of beryllium-7 on the component was assumed to be proportional to their surface area wetted by liquid lithium. As result, the most severe condition was that around the heat exchanger with surface area of 576 m. The dose equivalent rate was about 10 Sv/h, several orders higher than 10 Sv/h a limit considering ICRP recommendation. The dose rate can be reduced below the limit by a 22 cm-thick iron-shield or a 6.5 cm-thick lead-shield.
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
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.
Tanigawa, Hiroyasu; Sakasegawa, Hideo; Ogiwara, Hiroyuki*; Kishimoto, Hirotatsu*; Koyama, Akira*
Journal of Nuclear Materials, 367-370(1), p.132 - 136, 2007/08
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 (873K30k 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 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.
Nakamura, Hiroo; Ida, Mizuho; Chida, Teruo; Shiba, Kiyoyuki; Shimizu, Katsusuke*; Sugimoto, Masayoshi
Journal of Nuclear Materials, 367-370(2), p.1543 - 1548, 2007/08
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.
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
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 C is very small values. These data contribute to a part of materials database for ITER Test blanket design work.
Uwaba, Tomoyuki; Ukai, Shigeharu; Nakai, Tatsuyoshi*; Fujiwara, Masayuki*
Journal of Nuclear Materials, 367-370(2), p.1213 - 1217, 2007/08
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.
To, Kentaro*; Shikama, Tatsuo*; Nagata, Shinji*; Tsuchiya, Bun*; Yamauchi, Michinori; Nishitani, Takeo
Journal of Nuclear Materials, 367-370(2), p.1128 - 1132, 2007/08
no abstracts in English
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
A 300C 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 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 300C by Thermal Desorption Specroscopy (TDS). The observed mass number was 2, 17, 18, 28, 32, and 44. The number corresponds to hydrogen (H), ammonia (NH), water (HO), carbon monoxide (CO), oxygen (O), carbon dioxide (CO), respectively. The main outgas component from the resin at 100150C was NH and HO. The outgas of NH and HO from the resin have been measured, however, the neutron shielding performance of the resin after 200C 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 g/g at 300C. The neutron shielding performance of the developed resin at 300C 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 327C library and 20C library are almost same.