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

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

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.

Investigation of the chemical form of ruthenium compounds in the vitrification process, 4; RuO 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 crystal by reacting to raw materials for the vitrification process. In this study, the generation reaction to RuO was confirmed by heating Ru-La-Na mix nitrates and the raw materials.

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

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 470C by multi-ion beam irradiation, cavities are observed from 0.8 to 1.2m (damage level 20 dpa) from an irradiation surface at 470C irradiated specimen. On the other hand, few cavities are observed in same region at 400C irradiation. This result shows that a swelling peak temperature with F82H weldments by ion irradiation will be also around 470C.

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

Towards a practical realization of high-precision non-destructive assay of a spent nuclear fuel assembly; Understanding the accelerator requirements for using transmission nuclear resonance fluorescence

Angell, C.; Hajima, Ryoichi

no journal, , 

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 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.

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.