Monte Carlo Shielding Calculations for a Spent Fuel Transport Cask with Automated Monte Carlo Variance Reduction

• ASAMI Mitsufumi

  National Maritime Research Institute

• SAWAMURA Hidenori

  MHI Nuclear Engineering Co., Ltd

• NISHIMURA Kazuya

  MHI Nuclear Engineering Co., Ltd

抄録

For the purpose of performing reasonable shielding calculation of a spent fuel transport cask, the use of Monte Carlo methods has been proposed for solving the radiation transport problem on a detailed structure of the transport cask considering fixed neutron sources. A SMIRE (Simplified MCNP-ANISN_W Variance Reduction) system has been developed in the present study, which is possible to generate automatically the lower weight boundary of the weight window for each mesh based on the Consistent Adjoint Driven Importance Sampling (CADIS). Compared with the case of the importance based on the empirical formula, the figure of merit is increased by a factor of 25. In this system, it is possible to calculate the weight suitable for the distantly-positioned detector point from the fuel effective region to introduce the relaxation factor which relaxes the increase of the particle numbers at the boundary of weight window meshes that are generally caused by the large attenuation of adjoint flux. This system is used for a variety of the radiation transport problems as well as the transport cask.

収録刊行物

• Progress in Nuclear Science and Technology

  Progress in Nuclear Science and Technology 2 (0), 860-865, 2011-10-01

  Atomic Energy Society of Japan

キーワード

• CADIS methodology
• relaxation factor
• weight window
• figure of merit
• Monte Carlo shielding
• analysis
• empirical formula
• adjoint flux

詳細情報

• CRID: 1390569647113004288
• NII論文ID: 40019316942
• NII書誌ID: AA12785802
• DOI: 10.15669/pnst.2.860
• NDL書誌ID: 023761931
• ISSN: 21854823
• Web Site: https://ndlsearch.ndl.go.jp/books/R000000004-I023761931
• 本文言語コード: en
• データソース種別:
  • JaLC
  • NDL
  • Crossref
  • CiNii Articles
• 抄録ライセンスフラグ: 使用不可