380. MOX燃料炉物理試験MISTRALの解析

• 石井 一弥

  (株)日立製作所

• 巽 雅洋

  原子燃料工業(株)

• 菅 太郎

  三菱重工業(株)

• 安藤 良平

  (株)東芝

• 山本 徹

  (財)原子力発電技術機構

• 岩田 豊

  (財)原子力発電技術機構

• 馬野 琢也

  (財)原子力発電技術機構

• 神田 僚司

  (財)原子力発電技術機構

書誌事項

タイトル別名

• Analysis of MOX Core Physics Experiments MISTRAL

抄録

Nuclear Power Engineering Corporation (NUPEC) has been performing conceptual design studies of high moderation full MOX LWR cores that aim for increasing fissile Pu consumption rate and reducing residual Pu in discharged MOX fuel. As part of these studies, NUPEC, French Atomic Energy Commission (CEA) and their industrial partners implemented an experimental program, MISTRAL, that was devoted to measuring the core physics parameters of such advanced cores. The program consists of one reference UO₂ core, two homogeneous full MOX cores and one full MOX PWR mock-up core that have higher moderation ratio than the conventional lattice.<BR>NUPEC has been analyzing the experimental results with the diffusion and the transport calculations by the SRAC code system and the continuous energy Monte Carlo calculations by the MVP code with the common nuclear data file, JENDL-3.2. The calculation results well reproduce the experimental data approximately within the same range of the experimental uncertainty. This indicates that these applied analysis methods give the same accuracy for the UO₂ core and MOX cores, for the different moderation MOX cores, and for the homogeneous and the mock-up MOX cores.

収録刊行物

• 日本原子力学会和文論文誌

  日本原子力学会和文論文誌 2 (1), 39-54, 2003

  一般社団法人 日本原子力学会

キーワード

• MOX
• LWR type reactors
• core physics experiment
• MISTRAL
• high moderation
• experimental analysis
• SRAC
• MVP
• JENDL-3.2

詳細情報

• CRID: 1390001205186751616
• NII論文ID: 130003935903
• DOI: 10.3327/taesj2002.2.39
• COI: 1:CAS:528:DC%2BD3sXktVOls7g%3D
• ISSN: 21862931
• ISSN: 13472879
• データソース種別:
  • JaLC
  • Crossref
  • CiNii Articles
• 抄録ライセンスフラグ: 使用不可