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- TERASAKA Haruo
- Systems Safety Department Nuclear Power Engineering Corporation Toshiba Nuclear Engineering Lab.
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- MAKITA Atsushi
- Analysis Department Toshiba Advanced System Corporation
書誌事項
- タイトル別名
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- Numerical Analysis of the PHEBUS Contai
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This paper describes numerical analysis of the PHEBUS FP containment thermal-hydraulics. PHEBUS FP is an international project undertaken with the aim of evaluating the behavior of radioactive fission products released from a LWR pressure vessel into the containment vessel during a hypothetical severe accident. Six integral in-pile tests have been planned and are being carried out at Cadarache, France. The European Union, the United States, Canada, Korea and Japan are participating in this project. From Japan, the Nuclear Power Engineering Corporation and the Japan Atomic Energy Research Institute are collaborating the other parties involved in the project. <BR>Since the behavior of fission products is strongly dependent on the surrounding environmental conditions, accurate prediction of the thermal-hydraulics in the containment vessel is essential to accurately evaluate the behavior. Characteristics of condensation heat transfer in the presence of noncondensable gases play a key role in the PHEBUS thermal-hydraulics, especially under the condition of high noncondensable gas mass fraction. Many models for condensation heat transfer in the presence of noncondensable gases have been proposed. However, these models were not found suitable for PHEBUS analysis, because they were focused on the low noncondensable gas mass fraction condition.<BR>In this study, a single-phase multi-component code, TFLOW-FP has been newly developed to predict thermal-hydraulics in the PHEBUS FP containment. Moreover, a new degradation factor correlation for the condensation heat transfer coefficient due to the presence of noncondensable gases has also been developed and incorporated into the code. This code was applied for analysis of the thermal-hydraulic benchmark tests and the first in-pile test, FPTO. The results show that the code can predict the total pressure, gas temperature distributions, the relative humidity in the containment vessel and steam condensation rate on the surface of condenser rods very well.
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 (7), 666-678, 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
- 1390001204094023168
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- NII論文ID
- 10002077509
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- NII書誌ID
- AA00703720
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- COI
- 1:CAS:528:DyaK2sXls1KitLs%3D
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- ISSN
- 18811248
- 00223131
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- NDL書誌ID
- 4260275
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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