ガラス固化体からの放射線量に関する検討

• Material Type: 文書・図像類
• Title: ガラス固化体からの放射線量に関する検討
• Author/Editor: 若杉 圭一郎; 宮原 要; 牧野 仁史; 石黒 勝彦; 澤村 英範; 根山 敦史; 西村 和哉
• Author Heading: 若杉 圭一郎; 宮原 要; 牧野 仁史; 石黒 勝彦; 澤村 英範; 根山 敦史; 西村 和哉
• Publication Date: 2003-11
• Publication Date (W3CDTF): 2003-11
• Alternative Title: Study on Radiation Dose from Vitrified Waste
• Periodical title: JNC TN8400 2003-022
• Text Language Code: jpn
• Target Audience: 一般
• Note (General): 本検討では,第2次取りまとめのレファレンスケースで設定したガラス固化体と同様の仕様を想定したガラス固化体を対象に、キャニスタおよびオーバーパックによる遮へいの効果を考慮して、ガラス固化体からの放射線量を定量的に把握した。さらに,ガラス固化体からの放射線量の特徴を遮へいの観点から把握することを目的として,管理区域の基準値までガラス固化体からの放射線量を減衰させるために必要なコンクリート壁の厚さを評価した。; The radiation dose from the vitrified waste which is the same specification set in the Reference Case of the second progress report (H12 report) was evaluated quantitatively taking into account of the shield of the canister and the overpack. In order to understand the feature of radiation dose from the vitrified waste in terms of shielding, the thickness of the concrete shield to decrease less than safety standard for a radiation controlled zone was evaluated. Main results are summarized as follows. (1)The effective dose rates in the case considering the vitrified waste and the canister decrease approximate 4$\sim$5 orders of magnitude during the period of 1,000 years after vitrification due to decay of short half-life radionuclides. The effective dose rate doesn't decrease from 1,000 to 10,000 years. (2)The effective dose rates at the outside of overpack in the case considering the vitrified waste, canister and overpack are smaller than those inside of overpack approximate 5 orders of magnitude during the period of 100 years due to shielding effect of the overpack. However this difference is relatively small after 100 years since the contribution of $\gamma$ radiation to total effective dose rates decrease due to decay of fission products. (3)Excepting a few cases, the result using the old law (dose equivalent rate) is larger than the result using the new law (effective dose rate). However the difference between these results is less than factor of 1.2. (4)The thickness of the concrete shield required to attenuate the effective dose during the period of 50 years less than safety standard for a radiation controlled zone is calculated as approximate 0.8m$\sim$1.5m. The important factors to determine the thickness of the concrete shield are the $\gamma$ radiation in the case of vitrified waste and the canister, and the neutron radiation in the case of vitrified waste, canister and overpack.
• Source: /JNC-TN8400-2003-022.pdf （fulltext）; https://jopss.jaea.go.jp/pdfdata/JNC-TN8400-2003-022.pdf
• Access Restrictions: 限定公開
• Data Provider (Database): 国立情報学研究所 : 学術機関リポジトリデータベース（IRDB）（機関リポジトリ）; https://irdb.nii.ac.jp
• Original Data Provider (Database): 日本原子力研究開発機構 : JOPSS:JAEA Originated Papers Searching System

• Summary, etc.: 本検討では,第2次取りまとめのレファレンスケースで設定したガラス固化体と同様の仕様を想定したガラス固化体を対象に、キャニスタおよびオーバーパックによる遮へいの効果を考慮して、ガラス固化体からの放射線量を定量的に把握した。さらに,ガラス固化体からの放射線量の特徴を遮へいの観点から把握することを目的とし...