Experimental investigation of the impulse gas injection into liquid and the use of experimental data for verification of the HYDRA-IBRAE/LM thermohydraulic code
- Авторы: Lobanov P.D.1, Usov E.V.1, Butov A.A.1, Pribaturin N.A.1, Mosunova N.A.2, Strizhov V.F.2, Chukhno V.I.3, Kutlimetov A.E.3
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Учреждения:
- Novosibirsk Branch of the Nuclear Safety Institute
- Nuclear Safety Institute
- Novosibirsk National Research State University
- Выпуск: Том 64, № 10 (2017)
- Страницы: 770-776
- Раздел: Nuclear Power Stations
- URL: https://bakhtiniada.ru/0040-6015/article/view/172841
- DOI: https://doi.org/10.1134/S004060151710007X
- ID: 172841
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Аннотация
Experiments with impulse gas injection into model coolants, such as water or the Rose alloy, performed at the Novosibirsk Branch of the Nuclear Safety Institute, Russian Academy of Sciences, are described. The test facility and the experimental conditions are presented in details. The dependence of coolant pressure on the injected gas flow and the time of injection was determined. The purpose of these experiments was to verify the physical models of thermohydraulic codes for calculation of the processes that could occur during the rupture of tubes of a steam generator with heavy liquid metal coolant or during fuel rod failure in water-cooled reactors. The experimental results were used for verification of the HYDRA-IBRAE/LM system thermohydraulic code developed at the Nuclear Safety Institute, Russian Academy of Sciences. The models of gas bubble transportation in a vertical channel that are used in the code are described in detail. A two-phase flow pattern diagram and correlations for prediction of friction of bubbles and slugs as they float up in a vertical channel and of two-phase flow friction factor are presented. Based on the results of simulation of these experiments using the HYDRA-IBRAE/LM code, the arithmetic mean error in predicted pressures was calculated, and the predictions were analyzed considering the uncertainty in the input data, geometry of the test facility, and the error of the empirical correlation. The analysis revealed major factors having a considerable effect on the predictions. The recommendations are given on updating of the experimental results and improvement of the models used in the thermohydraulic code.
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Об авторах
P. Lobanov
Novosibirsk Branch of the Nuclear Safety Institute
Email: usovev@gmail.com
Россия, Novosibirsk, 630090
E. Usov
Novosibirsk Branch of the Nuclear Safety Institute
Автор, ответственный за переписку.
Email: usovev@gmail.com
Россия, Novosibirsk, 630090
A. Butov
Novosibirsk Branch of the Nuclear Safety Institute
Email: usovev@gmail.com
Россия, Novosibirsk, 630090
N. Pribaturin
Novosibirsk Branch of the Nuclear Safety Institute
Email: usovev@gmail.com
Россия, Novosibirsk, 630090
N. Mosunova
Nuclear Safety Institute
Email: usovev@gmail.com
Россия, Moscow, 115191
V. Strizhov
Nuclear Safety Institute
Email: usovev@gmail.com
Россия, Moscow, 115191
V. Chukhno
Novosibirsk National Research State University
Email: usovev@gmail.com
Россия, Novosibirsk, 630090
A. Kutlimetov
Novosibirsk National Research State University
Email: usovev@gmail.com
Россия, Novosibirsk, 630090
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