Progress of research and development of Full-W ITER divertor, 2; Deformation of tungsten after high heat flux testing for full-scale prototypes of plasma facing units

Seki, Yohji; Ezato, Koichiro; Suzuki, Satoshi; Yokoyama, Kenji; Yamada, Hirokazu; Hirayama, Tomoyuki

no journal, , 

JAEA is in charge of manufacturing a high heat flux (HHF) component, so-called "Outer vertical targets" in ITER divertor. The plasma facing units (PFUs) of the ITER divertor are subjected to HHF from plasma. The tungsten (W) armor tiles are metallurgically bonded onto cooling tubes made of copper alloy to achieve sufficient heat removal capability. Targets of R&D for full-W divertor are (1) to manufacture of full-scale PFU and (2) to demonstrate a durability of a heat removal capability. In the R&D through the manufacturing full-scale prototype PFU, JAEA demonstrated durability of heat removal capability of W armor tiles for HHF testing of 5000 10 MW/m and 1000 20 MW/m. After the HHF testing, gaps of about 0.5 mm between the W tiles decreased by about because the W tile was deformed by plastic strain. This paper reports deformation of shape on W surface in detail.

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.

Examination of the decreasing effect on internal exposure by sheltering, 2; Sensitivity analysis

Hirouchi, Jun; Takahara, Shogo; Watanabe, Masatoshi; Munakata, Masahiro

no journal, , 

no abstracts in English

Analyses of in-pile experiments by an analysis code on initiating phase of core disruptive accident in sodium cooled fast reactors,1; Summary of SAS4A code and CABRI in-pile experiments

Fukano, Yoshitaka; Imaizumi, Yuya; Yoshioka, Naonori*; Akahori, Hisashi*

no journal, , 

no abstracts in English

Analyses of in-pile experiments by an analysis code on initiating phase of core disruptive accident in sodium cooled fast reactors, 3 Analysis of LT4 test

Yoshioka, Naonori*; Fukano, Yoshitaka

no journal, , 

no abstracts in English

Development of non-destructive examination technique for fuel debris using X-ray computed tomography

Ishimi, Akihiro; Katsuyama, Kozo; Akasaka, Naoaki; Misawa, Susumu*

no journal, , 

no abstracts in English