Best Estimate BWR Transient with TRACG Assessment using Data from BWR Startup Tests and LOCA Integral Tests
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- UTSUNOI Hideaki
- Hitachi Works, Hitachi, Ltd.
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- SAKUMA Takayuki
- Hitachi Works, Hitachi, Ltd.
書誌事項
- タイトル別名
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- Best Estimate BWR Transient Analysis with TRACG Assessment using Data from BWR Startup Tests and LOCA Integral Tests.
- Best Estimate BWR Transient with TRACG
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TRACG is a new version of the best estimate BWR transient analysis code, which utilizes a multi-dimensional two-fluid model for the thermal hydraulics and a three-dimensional neutron kinetics model. A three-dimensional neutronics, a fully implicit integration scheme and models for advanced BWR components have been implemented in the code upon TRAC-BF1.<BR>Assessment of TRACG has been performed in this study for the predictive capability of plant transients, which include thermal-hydraulic and neutronic interactions, as affected by responses of the plant control system. Simulations were presented for BWR representative transient tests, which were done as part of a series of BWR5 startup tests. As for the capability to predict thermal hydraulics during the design basis LOCAs, simulations were presented for the LOCA integral tests conducted in the ROSA-III at JAERI and the Hitachi TBL, which had been used for assessment of the TRAC former version.<BR>Consequently, (1) the space-dependent power flow transitions in a BWR were confirmed by TRACG simulations in which the module coupled with neutronics and thermal hydraulics during transients has been newly introduced, and (2) the characteristic thermal-hydraulic phenomena including multi-channel effects during the design basis LOCAs were confirmed, as well as the TRAC former version, by TRACG simulations on which the influence due to a fully implicit integration scheme has not extended. Capability of TRACG to predict BWR transients ranging from simple plant operational transients to design basis LOCAs was successfully demonstrated.
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) 35 (8), 607-620, 1998
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
- 1390282679071610880
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- NII論文ID
- 10002079951
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- NII書誌ID
- AA00703720
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- COI
- 1:CAS:528:DyaK1cXmsVegsb0%3D
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- ISSN
- 18811248
- 00223131
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- NDL書誌ID
- 4541367
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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
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- 使用不可