Irradiation test about oxidation-resistant graphite in WWR-K research reactor
柴田 大受; 角田 淳弥; 坂場 成昭; 大崎 貴士*; 加藤 秀樹*; 井澤 祥一*; 武藤 剛範*; Gizatulin, S.*; Shaimerdenov, A.*; Dyussambayev, D.*; Chakrov, P.*
Shibata, Taiju; Sumita, Junya; Sakaba, Nariaki; Osaki, Takashi*; Kato, Hideki*; Izawa, Shoichi*; Muto, Takenori*; Gizatulin, S.*; Shaimerdenov, A.*; Dyussambayev, D.*; Chakrov, P.*
Graphite are used for the in-core components of HTGR, and it is desirable to enhance oxidation resistance to keep much safety margin. SiC coating is the candidate method for this purpose. JAEA and four Japanese graphite companies are studying to develop oxidation-resistant graphite. Neutron irradiation test was carried out by WWR-K reactor of INP of Kazakhstan through ISTC partner project. The total irradiation cycles of WWR-K operation was 10 cycles by 200 days. Irradiation temperature about 1473 K would be attained. The maximum fast neutron fluence (E $$>$$0.18 MeV) for the capsule irradiated at a central irradiation hole was preliminary calculated as 1.2$$times$$10$$^{25}$$/m$$^{-2}$$, and for the capsule at a peripheral irradiation hole as 4.2$$times$$10$$^{24}$$/m$$^{-2}$$. Dimension and weight of the irradiated specimens were measured, and outer surface of the specimens were observed by optical microscope. For the irradiated oxidation resistant graphite, out-of-pile oxidation test will be carried out at an experimental laboratory.
使用言語 : English
掲載資料名 : Proceedings of 8th International Topical Meeting on High Temperature Reactor Technology (HTR 2016) (CD-ROM)
ページ数 : p.567 - 571
発行年月 : 2016/11
出版社名 : American Nuclear Society
発表会議名 : 8th International Topical Meeting on High Temperature Reactor Technology (HTR 2016)
開催年月 : 2016/11
開催都市 : Las Vegas
開催国 : U. S. A.
特許データ :
論文URL :
キーワード : 高温ガス炉; 黒鉛; 耐酸化; 中性子; ISTC
使用施設 :
広報プレスリリース :
受委託・共同研究相手機関 : Institute of Nuclear Physics of National Nuclear Center, Kazakhstan
登録番号 : BB20161094
抄録集掲載番号 : 45000091
論文投稿番号 : 18427
Accesses  (From Jun. 2, 2014)
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