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- JEONG Jae-Jun
- Korea Atomic Energy Research Institute (KAERI)
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- SIM Suk Ku
- Korea Atomic Energy Research Institute (KAERI)
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- LEE Sang Yong
- Korea Power Engineering Company, Inc. (KOPEC)
書誌事項
- タイトル別名
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- Development and Assessment of the COBRA
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抄録
<BR>The COBRA/RELAP5 code has been developed to combine the realistic three-dimensional reactor vessel model of COBRA-TF with RELAP5/MOD3.2, thus to produce an advanced system analysis code with a multidimensional thermal-hydraulic module. This paper provides the integration scheme of the two codes. The results of developmental assesments are also provided, which include single channel tests, manometric flow oscillation problem, THTF Test 105, and LOFT L2-3 large break loss-of-coolant experiment. From the single channel tests, the integration scheme was proven to be valid. Other simulation results showed good agreement with the experimental data. The computational speed was also satisfactory. Therefore, the assessment confirmed that the COBRA/RELAP5 code can be a promising tool for analysis of complicated, multimensional, two-phase flow system transients.
収録刊行物
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- Journal of Nuclear Science and Technology(日本原子力学会英文論文誌)
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Journal of Nuclear Science and Technology(日本原子力学会英文論文誌) 34 (11), 1087-1098, 1997
一般社団法人 日本原子力学会
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詳細情報 詳細情報について
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- CRID
- 1390282679070693632
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- NII論文ID
- 10002078440
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- NII書誌ID
- AA00703720
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- COI
- 1:CAS:528:DyaK1cXhsVyktg%3D%3D
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- ISSN
- 18811248
- 00223131
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- NDL書誌ID
- 4346186
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- 本文言語コード
- en
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- データソース種別
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- JaLC
- NDL
- Crossref
- CiNii Articles
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- 抄録ライセンスフラグ
- 使用不可