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Oral presentation

Study on rapid analysis of the radioactive tritium in groundwater into the Fukushima-1 Nuclear Power Plant, 1; Applicability confirmation of the tritium column

Sasaki, Takayuki*; Akimoto, Yuji*; Seki, Kotaro; Nagano, Misato*; Ishimori, Kenichiro; Ueno, Takashi; Kameo, Yutaka

no journal, , 

no abstracts in English

Oral presentation

Uncertainty of criticality analysis of UO$$_{2}$$-concrete system under randomization

Ueki, Taro

no journal, , 

Analysis framework under indeterminate material distribution is investigated for the Monte Carlo (MC) criticality calculation of continuously mixed media formed via molten core concrete interaction. Randomized Weierstrass functions (RWF) are utilized to represent the volume fractions of constituent materials. The possibility of several percent fluctuation of effective multiplication factor is shown by the MC simulation with delta-tracking.

Oral presentation

Remote technology development for function advancement of research base, 7; Feasibility study of a portable Compton camera for visualizing radioactive substances

Sato, Yuki; Kishimoto, Aya*; Kaburagi, Masaaki; Kataoka, Jun*; Torii, Tatsuo

no journal, , 

no abstracts in English

Oral presentation

Effect of hydrocarbons on tritium oxidation reactor for ITER detritiation system

Edao, Yuki; Iwai, Yasunori; Sato, Katsumi; Hayashi, Takumi

no journal, , 

no abstracts in English

Oral presentation

Development of active neutron NDA techniques for nuclear non-proliferation, 3; Experimental evaluation of neutron flux distribution in the DDA system

Maeda, Makoto; Komeda, Masao; Tobita, Hiroshi; Ozu, Akira; Kureta, Masatoshi

no journal, , 

JAEA has started to develop a technology which can be applicable to high radioactive special nuclear materials such as next-generation fuel cycle products. We have been developed Non-destructive assay system Active-N as a test equipment which utilizes D-T neutron generator. In a system for Differential Die-Away (DDA) method which is tested in Active-N, it is important to evaluate neutron flux to check the performance of the system. In this research, we have evaluated neutron flux in a system for Fast Neutron Direct Interrogation method which is a kind of DDA method by activation method and Monte Carlo simulation by using PHITS.

Oral presentation

The Supports for local governments used the walking survey

Terunuma, Hirotaka; Tanaka, Kiwamu; Kabumoto, Hiroshi; Haginoya, Masashi; Sano, Naruto; Takahashi, Masatomi; Hoshino, Masato; Aoki, Isao; Asazuma, Shinichiro

no journal, , 

no abstracts in English

Oral presentation

Investigation of the chemical form of ruthenium compounds in the vitrification process, 4; RuO$$_{2}$$ generation by reaction with Ru-La-Na mix nitrates and raw materials for vitrification

Nagai, Takayuki; Kobayashi, Hidekazu; Okamoto, Yoshihiro; Sato, Nobuaki*; Inose, Takehiko*; Sato, Seiichi*; Hatakeyama, Kiyoshi*; Seki, Katsumi*

no journal, , 

It is thought that a generated ruthenium compound grows from a high level radioactive liquid waste into RuO$$_{2}$$ crystal by reacting to raw materials for the vitrification process. In this study, the generation reaction to RuO$$_{2}$$ was confirmed by heating Ru-La-Na mix nitrates and the raw materials.

Oral presentation

Development of active neutron NDA techniques for nuclear non-proliferation, 4; Design study on NRTA technique for quantification of isotopes of nuclear materials

Tsuchiya, Harufumi; Kitatani, Fumito; Kureta, Masatoshi; Maeda, Makoto

no journal, , 

no abstracts in English

Oral presentation

Development of active neutron NDA techniques for nuclear non-proliferation, 2; Design study on DDA technique for nuclear materials

Ozu, Akira; Maeda, Makoto; Komeda, Masao; Tobita, Hiroshi; Kureta, Masatoshi

no journal, , 

no abstracts in English

Oral presentation

Oral presentation

Void swelling behavior of multi-ion irradiated F82H

Ando, Masami; Tanigawa, Hiroyasu; Kurotaki, Hironori

no journal, , 

Reduced activation ferritic/martensitic steel (RAFM) is the most promising candidate for the blanket structural material in DEMO fusion reactor. Void swelling behavior as well as mechanical properties change under high dose fusion neutron irradiation is very important issue on R&D of RAFM steels (F82H). The purposes of this study are to investigate the void swelling behavior with F82H IEA, Mod-3 and BA07 heats after multi-ion-irradiation and to confirm the peak swelling temperature. From results of microstructure observation for F82H Mod-3 and BA07 heats at 470$$^{circ}$$C by multi-ion beam irradiation, cavities are observed from 0.8 to 1.2$$mu$$m (damage level $$sim$$20 dpa) from an irradiation surface at 470$$^{circ}$$C irradiated specimen. On the other hand, few cavities are observed in same region at 400$$^{circ}$$C irradiation. This result shows that a swelling peak temperature with F82H weldments by ion irradiation will be also around 470$$^{circ}$$C.

Oral presentation

Tensile fracture behavior of high dose irradiated reduced activation ferritic/martensitic steel F82H

Tanigawa, Hiroyasu; Sakasegawa, Hideo; Hirose, Takanori; Kato, Yutai*

no journal, , 

Reduced activation ferritic/martensitic steel, F82H, is the first candidate DEMO blanket structural material, and the preparation of high dose irradiation database is underway. 87 dpa irradiation was completed and post irradiation tensile tests were conducted at the irradiation temperature. The results of the tensile test and fractography on 300$$^{circ}$$C irradiated specimens will be reported.

Oral presentation

Study on quantity fabrication and purity control of lithium-lead as fusion liquid breeder

Park, C.-H.; Kondo, Masatoshi*; Nozawa, Takashi; Tanigawa, Hiroyasu

no journal, , 

no abstracts in English

Oral presentation

Remote technology development for function advancement of research base, 8; Laser diagnostics technology development for characterization of radioactivated concrete

Yamada, Tomonori; Daido, Hiroyuki; Kawamura, Hiroshi; Shimada, Yoshinori*

no journal, , 

no abstracts in English

Oral presentation

2-d scintillation detector for neutron protein diffractometer in J-PARC, 2; FPGA-based center of gravity circuit for a photon counting detector

Ebine, Masumi; Nakamura, Tatsuya; To, Kentaro; Honda, Katsunori; Birumachi, Atsushi; Sakasai, Kaoru

no journal, , 

The center of gravity method is developed for a two-dimensional scintillator neutron detector for a protein diffractometer in the J-PARC/MLF. Details of the FPGA-based digital signal processing circuit and the detector performances are presented.

Oral presentation

Applicability study of Friction-Stir Process on CuCrZr-alloy

Hamaguchi, Dai; Morisada, Yoshiaki*; Fujii, Hidetoshi*; Tanigawa, Hiroyasu; Ozawa, Kazumi

no journal, , 

Friction-Stir Process (FSP) were applied to OFCCu and ITER-Gr CuCrZr cooling pipe in order to evaluate the applicability of the process to mechanically strengthening pure-Cu and possible structural improvement on CuCrZr alloy. Former study revealed that the achievement of very fine grain structure at certain rotational speed and also the hardness increase to about 1.5 time higher within the stir-zone but also relatively shallow stir-zone depth compared to a length of the pin and rapid void formations. In order to avoid this phenomenon, compulsory cooling with liquid Co$$_{2}$$ during the process was introduced, and the results showed the absence of void formation at relatively lower rational speed of around 300 RPM and also maintaining the effective stir-zone to relatively higher rotational speed compared to non-cooling tests. Similar FSP tests with compulsory cooling were also applied to ITER-Gr CuCrZr cooling pipe for divertor application. The detail will be reported on the presentation.

Oral presentation

Simulation of TVF glass melter operation with forming of thick sediments composed of platinum group elements

Asahi, Yoshimitsu; Nakajima, Masayoshi; Ayame, Yasuo

no journal, , 

no abstracts in English

Oral presentation

Probabilistic risk assessment method development for high temperature gas-cooled reactors

Sato, Hiroyuki; Nishida, Akemi; Furuya, Osamu*; Muramatsu, Ken*; Itoi, Tatsuya*; Takada, Tsuyoshi*; Tanabe, Masayuki*; Yamamoto, Tsuyoshi*

no journal, , 

The proposed research aims to establish a probabilistic risk assessment method for high temperature gas-cooled reactors fully utilizing their design and safety characteristics. The method will be developed for the incorporation of a graded approach as well as a component failure evaluation model using the operation and maintenance experience in the high temperature engineering test reactor into an accident frequency analysis. In addition, a source term evaluation method considering failures in core graphite components will be developed.

Oral presentation

Present status and future prospects of the MVP code

Nagaya, Yasunobu

no journal, , 

no abstracts in English

249 (Records 1-20 displayed on this page)