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Journal Articles

Experimental validation of tensile properties measured with thick samples taken from MEGAPIE target

Saito, Shigeru; Suzuki, Kazuhiro; Hatakeyama, Yuichi; Suzuki, Miho; Dai, Y.*

Journal of Nuclear Materials, 534, p.152146_1 - 152146_16, 2020/06

 Times Cited Count:1 Percentile:39.17(Materials Science, Multidisciplinary)

A post-irradiation examination (PIE) was performed on the tensile specimens prepared from the MEGAPIE (MEGAwatt Pilot Experiment) target which were irradiated in flowing lead-bismuth eutectic (LBE). Thicknesses of the specimens were over two times larger than that of the standard specimen. The PIE revealed that the T91 specimens showed a 1.5-2.0 times larger total elongation (TE) compared to the literature values for a specimen with standard t/w (ratio of thickness to width). It could be suggested that the t/w and TE were strongly correlated. Then, we tried to investigate the effects of the t/w on the TE by comparing unirradiated specimens. We found that there was no t/w dependence on the strength and uniform elongation. On the other hand, the TE increases with increasing t/w. Based on the experimental data, we correlated the TE with various specimens t/w to estimate appropriate TE values, including that for the standard specimen.

Journal Articles

Current status of the control system for J-PARC accelerator complex

Yoshikawa, Hiroshi; Sakaki, Hironao; Sako, Hiroyuki; Takahashi, Hiroki; Shen, G.; Kato, Yuko; Ito, Yuichi; Ikeda, Hiroshi*; Ishiyama, Tatsuya*; Tsuchiya, Hitoshi*; et al.

Proceedings of International Conference on Accelerator and Large Experimental Physics Control Systems (ICALEPCS '07) (CD-ROM), p.62 - 64, 2007/10

J-PARC is a large scale facility of the proton accelerators for the multi-purpose of scientific researches in Japan. This facility consists of three accelerators and three experimental stations. Now, J-PARC is under construction, and LINAC is operated for one year, 3GeV synchrotron has just started the commissioning in this October the 1st. The completion of this facility will be next summer. The control system of accelerators established fundamental performance for the initial commissioning. The most important requirement to the control system of this facility is to minimize the activation of accelerator devices. In this paper, we show that the performances of each layer of this control system have been achieved in the initial stage.

JAEA Reports

Post-irradiation examinations of inert matrix nitride fuel irradiated in JMTR (01F-51A capsule)

Iwai, Takashi; Nakajima, Kunihisa; Kikuchi, Hironobu; Honda, Junichi; Hatakeyama, Yuichi; Ono, Katsuto; Matsui, Hiroki; Arai, Yasuo

JAEA-Research 2007-026, 75 Pages, 2007/03

JAEA-Research-2007-026.pdf:13.6MB

A plutonium nitiride fuel pin containing inert matrix such as ZrN and TiN was encapsuled in 01F-51A and irradiated in JMTR. Minor actinides are surrogated by plutonium. Average linear powers and burnups were 408W/cm, 30000MWd/t(Zr+Pu)(132000MWd/t-Pu) for (Zr,Pu)N and 355W/cm, 38000MWd/t(Ti+Pu)(153000MWd/t-Pu) for (TiN,PuN). The irradiated capsule was transported to Reactor Fuel Examination Facility and subjected to non-destructive and destructive post irradiation examinations. Any failure was not observed in theirradiated fuel pin. Very low fission gas release rate of about 1.6% was measured. The inner surface of cladding tube did not show any signs of chemical interaction with fuel pellets.

JAEA Reports

Development of fission gas measurement technique in the irradiated fuel pellet

Hatakeyama, Yuichi; Sudo, Kenji; Kanazawa, Hiroyuki

JAERI-Tech 2004-033, 29 Pages, 2004/03

JAERI-Tech-2004-033.pdf:1.65MB

The amount of fission gas (Kr, Xe) in irradiated fuel pellet increases with extending the burn up and that exerts a serious influence upon thermal and mechanical properties of light water reactor fuel. Therefore, the accumulation of the data on the release behavior of fission gas is important in the investigation program of safety and reliability for extended burn up fuel. In the post irradiation examination at the Reactor Fuel Examination Facility in JAERI,the fission gas which released into the plenum region from UO$$_{2}$$ pellet during irradiation has been measured by puncturing test of irradiated fuel rod. The results of puncturing test show the most of fission gas remained in the pellet. It can be seen that the additional release of fission gas might occur under higher burn up and accident conditions. To know the fission gas release behavior from irradiated fuel, the Out Gas analyzer(OGA)which has the performance to heat up the UO$$_{2}$$ pellet stepwise up to 2300$$^{circ}$$C and to measure the released fission gas instantly from the pellet has been developed and installed at RFEF.

Journal Articles

Application of hydrogen analysis by neutron imaging plate method to Zircaloy cladding tubes

Yasuda, Ryo; Nakata, Masahito; Matsubayashi, Masahito; Harada, Katsuya; Hatakeyama, Yuichi; Amano, Hidetoshi

Journal of Nuclear Materials, 320(3), p.223 - 230, 2003/08

 Times Cited Count:14 Percentile:68.9(Materials Science, Multidisciplinary)

Neutron radiography is one of effective tools to determine hydrided region in Zircaloy cladding tubes. In this work, the practicability of the neutron radiography for hydrogen analysis is further investigated by using standard samples with known hydrogen concentration. Local hydrogen concentration in hydrided Zircaloy tube is quantitatively estimated using the standard samples by neutron imaging plate (NIP) method. The local area is equivalent to a picture element in the image; e.g., 0.1mm$$times$$0.1mm. In addition, contribution of an oxide film in the tubes to the images is investigated using oxidized samples with hydrides or no hydride. In NIP images of oxidized tube no oxide film was recognized. Numerical image analysis also shows no effect of the oxide film on the image. These results show that the influence of oxygen on image contrast can be neglected when hydrogen analysis is performed on the Zircaloy tube with oxide film and hydrides by NIP method.

Journal Articles

Analysis of hydrogen content and distribution in hydrogen storage alloys using neutron radiography

Sakaguchi, Hiroki*; Hatakeyama, Keisuke*; Satake, Yuichi*; Fujine, Shigenori*; Yoneda, Kenji*; Kanda, Keiji*; Matsubayashi, Masahito; Esaka, Takao*

Kashika Joho Gakkai-Shi, 20(suppl.1), p.373 - 374, 2000/07

no abstracts in English

Oral presentation

Post irradiation examination on particle dispersed Uranium Rock-like Oxide (U-ROX) fuel, 3; Ceramographic examination

Shirasu, Noriko; Kuramoto, Kenichi*; Yamashita, Toshiyuki; Honda, Junichi; Hatakeyama, Yuichi; Ichise, Kenichi

no journal, , 

no abstracts in English

Oral presentation

Post-irradiation destructive examinations of inert matrix nitride fuels

Matsui, Hiroki; Iwai, Takashi; Honda, Junichi; Hatakeyama, Yuichi; Mita, Naoaki; Arai, Yasuo

no journal, , 

no abstracts in English

Oral presentation

New cross-over project new engineering to control materials behaviour at high energy and high fluence irradiation, 6; Crystal lattice strain changes in irradiated fuel

Nakamura, Jinichi; Amaya, Masaki; Honda, Junichi; Hatakeyama, Yuichi; Fuketa, Toyoshi; Kinoshita, Motoyasu*

no journal, , 

Based on the micro-X-ray diffraction results, lattice parameters and diffraction peak broadenings (increase of FWHM of diffraction peak) of high burnup UO$$_{2}$$ pellets were measured. Crystallite diameter, which corresponds to subdivided grain diameter, and strain distribution between crystallites can be evaluated from the increase of FWHM of diffraction peak by using Willamson-Hall method. Basesd on these data, the crystal microstructural change during irradiation was examined and the effects of irradiation for microstructure.

Oral presentation

Behavior of high burnup fuels under Reactivity-Initiated Accident (RIA) and Loss-of-Coolant Accident (LOCA) conditions, 6; Fracture conditions of fuel rods in a LOCA

Nagase, Fumihisa; Chuto, Toshinori; Hatakeyama, Yuichi; Fuketa, Toyoshi

no journal, , 

LOCA-simulated experiments were conducted with irradiated PWR and BWR fuel cladding ($$<$$77 GWd/t) to investigate high burnup effect on LWR fuel behavior under LOCA conditions. In the experiments, MDA, NDA, ZIRLO, M5 and Zircaloy-2 cladding specimens were sampled from the high burnup fuels irradiated at European power plants, and they were oxidized at 1563 to 1480 K and quenched. The cladding specimens oxidized to 18.3 to 27.3% ECR did not fracture during the quench. Consequently, the fracture boundary is not significantly reduced in the examined burnup range. Obvious high burnup effect was not seen also in the ballooning, rupture, fracture behavior.

Oral presentation

Post irradiation examination results of hydride neutron absorber for fast reactor, 2; Weight measurement, X-ray diffraction

Harada, Akio; Hatakeyama, Yuichi; Honda, Junichi; Matsui, Hiroki; Kurosaki, Ken*; Konashi, Kenji*

no journal, , 

no abstracts in English

Oral presentation

Research for the improvement of SCC evaluation method including the effect of radiation on water, 3; Influence of materials and dissolved oxygen related to oxide film structures on stainless steel under $$gamma$$-ray irradiation in high temperature water

Yamamoto, Masahiro; Nakahara, Yukio; Kato, Chiaki; Tsukada, Takashi; Suzuki, Kazuhiro; Hatakeyama, Yuichi; Watanabe, Atsushi*; Fuse, Motomasa*

no journal, , 

It is well known that the decomposition of water by irradiation results in the formation of oxidants, such as hydrogen peroxide (H$$_{2}$$O$$_{2}$$) and radical species. The oxidant such as H$$_{2}$$O$$_{2}$$ promote nobler electrochemical corrosion potential (ECP) of stainless steels (SSs) and change oxide instructors as corrosion products on stainless steels surface. In this work, the effects of $$gamma$$-ray irradiation and structure on the corrosion of Type 316L SS and 304L SS in high-temperature water of 288C were examined. We also consider structures of the oxide film on the stainless steel using transmission electron microscope (TEM) and Raman spectroscope in order to show the differences made by the irradiation and structure. In addition, the effect of diffusion restriction and surface reactions may become more significant Inside crevices, so that the environments are expected to be different from those in the outsides.

Oral presentation

Design and manufacturing of a small VDE-Free tokamak

Hatakeyama, Shoichi*; Tsutsui, Hiroaki*; Iio, Shunji*; Shimada, Ryuichi*; Shibata, Yoshihide; Ono, Noriyasu*; Akiyama, Tsuyoshi*; Suzuki, Yasuhiro*; Watanabe, Kiyomasa*

no journal, , 

no abstracts in English

Oral presentation

Self-irradiation effect on thermal conductivity of Zr$$_{0.70}$$Pu$$_{0.25}$$Cm$$_{0.05}$$N

Nishi, Tsuyoshi; Hayashi, Hirokazu; Hatakeyama, Yuichi; Kurata, Masaki

no journal, , 

no abstracts in English

Oral presentation

Post irradiation examination of the MEGAPIE samples at JAEA, 2

Saito, Shigeru; Kikuchi, Kenji*; Suzuki, Kazuhiro; Hatakeyama, Yuichi; Endo, Shinya; Suzuki, Miho; Okubo, Nariaki; Kondo, Keietsu

no journal, , 

The world's first megawatt-class lead-bismuth target, MEGAPIE (MEGAwatt Pilot Experiment), was dismantled and post irradiation examination (PIE) samples were prepared at PSI hot-lab. The samples were shipped to each institutions including JAEA. The samples were cut from the beam window (BW, T91) and the flow guide tube (FGT, SS316L). And all samples are prepared without LBE. The irradiation conditions of the specimens irradiated at SINQ target were as follows: proton energy was 580 MeV, irradiation temperatures were ranged from 251 to 341$$^{circ}$$C, and displacement damage levels were ranged from 0.16 to 1.57 dpa. PIE including SP (small punch) and three point bending tests were performed. SP tests were executed for T91 and SS316L specimens at R.T. in air condition. Specimen size for SP test with 2.4 mm steel-ball is 8 mm $$times$$ 8 mm $$times$$ 0.5 mm. T91 specimens were cut from the Spitze (triangle) sample and polished to thickness of 0.5 mm. The OM/SP specimens of SS316L were polished to thickness of 0.5 mm. Three point bending tests were executed for SS316L specimens at R.T. in air condition. The bend bar specimens of SS316L without notch were employed. Results of the SP tests and three point bending tests on the irradiated specimens will be presented at the workshop. Cross sectional observation on the Spitze sample and microstructural observation by TEM will be also reported.

Oral presentation

PSI SINQ specimen PIE at JAEA-WASTEF

Saito, Shigeru; Okubo, Nariaki; Endo, Shinya; Suzuki, Kazuhiro; Hatakeyama, Yuichi; Kikuchi, Kenji*

no journal, , 

To evaluate the lifetime of the beam window of an accelerator-driven transmutation system (ADS) and spallation neutron source, material irradiation programs, the STIP (SINQ target irradiation program, SINQ; Swiss spallation neutron source) and MEGAPIE (MEGAwatt Pilot Experiment), have been executed. Part of the specimens were transported to JAEA and post irradiation examination (PIE) of the specimens was carried out. In this presentation, in addition to representative results of the PIE, our experience and knowledge of irradiation experiments and PIE processes will be introduced. These information will be useful for high energy particle irradiation experiments and PIE planed under the frame work of RaDIATE.

16 (Records 1-16 displayed on this page)
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