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Journal Articles

Behavior of tritium in the vacuum vessel of JT-60U

Kobayashi, Kazuhiro; Torikai, Yuji*; Saito, Makiko; Alimov, V. Kh.*; Miya, Naoyuki; Ikeda, Yoshitaka

Fusion Science and Technology, 67(2), p.428 - 431, 2015/03

 Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)

Disassembly of the JT-60U torus was started in 2010 after 18 years deuterium operations. In the disassembly of the JT-60U torus, tritium retention in the vacuum vessel of the JT-60U is one of the most important safety issues for the fusion reactor. It was very important to study the tritium behavior in Inconel 625 from viewpoint of the clearance procedure in the future plan. After the tritium release for about 1 year at 298 K, the residual tritium in the specimen was released by heating up to 1073 K, and then the residual tritium in the specimen was measured by chemical etching method. Most of the chemical form of the released tritium was HTO. The contaminated specimen by tritium was released continuously the diffusible tritium under the ambient condition. In the tritium release experiment, most of tritium in the specimen was released during 1 year.

Journal Articles

Preliminary assessment for dust contamination of ITER in-vessel transporter

Saito, Makiko; Ueno, Kenichi; Maruyama, Takahito; Murakami, Shin; Takeda, Nobukazu; Kakudate, Satoshi; Nakahira, Masataka*; Tesini, A.*

Fusion Engineering and Design, 89(9-10), p.2352 - 2356, 2014/10

 Times Cited Count:7 Percentile:55.77(Nuclear Science & Technology)

After plasma operation of the ITER reactor, irradiated radioactive dust will accumulate in the vacuum vessel (VV). The In Vessel Transporter (IVT) will be installed in the VV and remove the blanket modules for maintenance. The IVT will be carried back to the Hot Cell Facilities (HCF) after exchanging the blanket, and the IVT itself also needs maintenance. It is considered that the maintenance workers will be exposed to the irradiated radioactive dust attached to the IVT surface. In this study, dust contamination of the IVT is evaluated to assess exposure during maintenance work in the HCF. The IVT contamination scenario is assumed in the ITER project. From plasma shut down until maintenance is performed on the IVT will take 345 days under the ITER project assumption. Under this scenario, the effective dose rate from irradiated radioactive dust was calculated as an infinite plate for each nuclide. As a result, W-181 and Ta-182 were the dominant nuclides for the effective dose rate. If all dust is W-181 or Ta-182, the effective dose rate is about 400 $$mu$$Sv/h and 100 $$mu$$Sv/h respectively. Nevertheless, using the dose limit determined by the ITER project and the estimated maximum maintenance time, the effective dose rate limit was calculated to be 4.18 $$mu$$Sv/h under these limited conditions. To satisfy the dose rate limit, decontamination processes were assumed and the dose rate after decontamination was evaluated.

Journal Articles

Dust removal experiments for ITER blanket remote handling system

Ueno, Kenichi; Aburadani, Atsushi; Saito, Makiko; Maruyama, Takahito; Takeda, Nobukazu; Murakami, Shin; Kakudate, Satoshi

Plasma and Fusion Research (Internet), 9, p.1405012_1 - 1405012_4, 2014/02

Journal Articles

Tritium distribution on the tungsten surface exposed to deuterium plasma and then to tritium gas

Isobe, Kanetsugu; Alimov, V. Kh.*; Taguchi, Akira*; Saito, Makiko; Torikai, Yuji*; Hatano, Yuji*; Yamanishi, Toshihiko

Journal of Plasma and Fusion Research SERIES, Vol.10, p.81 - 84, 2013/02

The distribution of hydrogen trapping sites on W surface exposed with D plasma was examined by the techniques of imaging plate and autoradiography. Recrystallized W specimens were exposed with D plasma at around 495 and 550 K to the same fluence of 10$$^{26}$$ D/m$$^{2}$$. Then, tritium was introduced into specimen by the exposure to tritium gaseous at 473 K. After that, the tritium distribution on W surface was examined by the techniques of imaging plate and autoradiography. From the results of the imaging plate, tritium was found to be highly concentrated within the area exposed with D plasma and the concentration of tritium was slightly varied even in that area. In the autoradiograph of W surface, it was found that tritium concentrated on the grain boundary and blisters.

Journal Articles

The H-Invitational Database (H-InvDB); A Comprehensive annotation resource for human genes and transcripts

Yamasaki, Chisato*; Murakami, Katsuhiko*; Fujii, Yasuyuki*; Sato, Yoshiharu*; Harada, Erimi*; Takeda, Junichi*; Taniya, Takayuki*; Sakate, Ryuichi*; Kikugawa, Shingo*; Shimada, Makoto*; et al.

Nucleic Acids Research, 36(Database), p.D793 - D799, 2008/01

 Times Cited Count:51 Percentile:74.53(Biochemistry & Molecular Biology)

Here we report the new features and improvements in our latest release of the H-Invitational Database, a comprehensive annotation resource for human genes and transcripts. H-InvDB, originally developed as an integrated database of the human transcriptome based on extensive annotation of large sets of fulllength cDNA (FLcDNA) clones, now provides annotation for 120 558 human mRNAs extracted from the International Nucleotide Sequence Databases (INSD), in addition to 54 978 human FLcDNAs, in the latest release H-InvDB. We mapped those human transcripts onto the human genome sequences (NCBI build 36.1) and determined 34 699 human gene clusters, which could define 34 057 protein-coding and 642 non-protein-coding loci; 858 transcribed loci overlapped with predicted pseudogenes.

Oral presentation

Decontamination assessment for ITER blanket remote handling system

Saito, Makiko; Maruyama, Takahito; Ueno, Kenichi; Takeda, Nobukazu; Kakudate, Satoshi

no journal, , 

In ITER, after plasma operation, The Blanket Remote Handling System (BRHS) will be installed in the vacuum vessel and it will remove and install the shield blanket module. BRHS will undergo hands-on maintenance in the maintenance area after the exchange of the shield blanket module. Since BRHS will be contaminated the radioactive dust in the vacuum vessel, the workers will be exposed by radioactive dust. In this study, potential contaminated areas and their respective dose rates from the BRHS using MCNP5 code to assess the exposure of maintenance workers. The assessment was performed using 3 types of equipment, vehicle manipulator, combination of cable handling and rail support, and sliding beam, which are installed in vacuum vessel or port. The dose calculations used the nuclides Ta-182 and W-181 and the dose was calculated from each of the 20 points spaced evenly around the equipment. As a result, there are some local points with high dose rates, which are exceed the target of acceptable dose limit for hands-on work in ITER (5 $$mu$$Sv/h) in vehicle manipulator and combination of cable handling and rail support. To decrease the dose rate, lead blocks were used for shielding and as a result, the dose rate decreased to around 2.5 $$mu$$Sv/h using 5 mm and 10 mm lead shielding.

Oral presentation

Analysis of radioactivation of dismantling components on JT-60U

Saito, Makiko; Sukegawa, Atsuhiko; Kobayashi, Kazuhiro; Miya, Naoyuki; Ikeda, Yoshitaka

no journal, , 

no abstracts in English

Oral presentation

Development of radiation hard components for ITER blanket remote handling system

Saito, Makiko; Anzai, Katsunori; Maruyama, Takahito; Noguchi, Yuto; Takeda, Nobukazu; Kakudate, Satoshi

no journal, , 

no abstracts in English

Oral presentation

Development of radiation hard components for ITER remote maintenance robot

Komai, Masafumi; Anzai, Katsunori; Noguchi, Yuto; Saito, Makiko; Maruyama, Takahito; Takeda, Nobukazu; Kakudate, Satoshi

no journal, , 

$$gamma$$-ray Environment of ~250 Gy/hr in International Thermonuclear Experimental Reactor (ITER) requires a full remote maintenance of in-vessel components such as blanket modules. JAEA is carrying out development of radiation hard components for the ITER blanket remote handling system. The current status of the radiation hardness component development is presented.

Oral presentation

Activity report of the emergency monitoring committee

Hosoda, Masahiro*; Saito, Kimiaki; Mikami, Satoshi; Sanada, Tetsuya*; Omori, Yasutaka*; Takeda, Hikaru*; Yamada, Takahiro*; Hirao, Shigekazu*; Tani, Kotaro*; Orita, Makiko*; et al.

no journal, , 

no abstracts in English

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