Thermal Conductivity Measurements on UO2+x from 300 to 1,400 K.

• AMAYA Masaki

  Nippon Nuclear Fuel Development Co., Ltd.

• KUBO Toshio

  Nippon Nuclear Fuel Development Co., Ltd.

• KOREI Yoshiaki

  Nippon Nuclear Fuel Development Co., Ltd.

書誌事項

タイトル別名

• Thermal Conductivity Measurements on UO

抄録

In order to evaluate the thermal conductivity of oxidized fuel pellets of leaker fuel rods, UO_2+x samples with x between 0.00 and 0.20 were prepared by an oxidation or a sintering method. Sample thermal diffusivities were measured by using a laser flash method from 300 to 1, 400K and their thermal conductivities were evaluated from multiplying the thermal diffusivities by the sample densities and the specific heat capacities derived from the literature. The thermal conductivities of UO_2+x were decreased with increasing hyperstoichiometry and they were expressed as a function of their hyperstoichiometry using the concentration of U₅₊ formed with the excess interstitial oxygen atoms.

収録刊行物

• Ｊｏｕｒｎａｌ　ｏｆ　Ｎｕｃｌｅａｒ　Ｓｃｉｅｎｃｅ　ａｎｄ　Ｔｅｃｈｎｏｌｏｇｙ（日本原子力学会英文論文誌）

  Ｊｏｕｒｎａｌ　ｏｆ　Ｎｕｃｌｅａｒ　Ｓｃｉｅｎｃｅ　ａｎｄ　Ｔｅｃｈｎｏｌｏｇｙ（日本原子力学会英文論文誌） 33 (8), 636-640, 1996

  一般社団法人 日本原子力学会

キーワード

• thermodynamic properties
• thermal conductivity
• thermal diffusivity
• laser flash method
• uranium oxides
• O/U ratios
• fuel pellets
• nuclear fuels
• temperature dependence
• thermal conductivity equation
• temperature range 0273-0400K
• temperature range 0400-1000K
• temperature range 1000-4000K

詳細情報

• CRID: 1390001204093732352
• NII論文ID: 10002075265
• NII書誌ID: AA00703720
• DOI: 10.3327/jnst.33.636
• COI: 1:CAS:528:DyaK28XmtVGmtLg%3D
• ISSN: 18811248
• ISSN: 00223131
• NDL書誌ID: 4060417
• Web Site: https://ndlsearch.ndl.go.jp/books/R000000004-I4060417
• 本文言語コード: en
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