نوع مقاله : مقاله فنی
نویسندگان
1 دانشکده علوم و فناوری‌های نوین، دانشگاه اصفهان، صندوق پستی: 73441-81746، اصفهان ـ ایران
2 پژوهشگاه علوم و فنون هسته‌ای، صندوق پستی: 1589-81465، اصفهان ـ ایران
3 دانشکده علوم، دانشگاه اصفهان، صندوق پستی: 73441-81746، اصفهان ـ ایران
چکیده
شبکه‌های نگه‌دارنده در مجتمع‌های سوخت هسته‌ای از جمله تجهیزاتی هستند که با ایجاد اغتشاش و هدایت جریان سیال در زیر کانال‌ها روی توزیع جریان سیال خنک‌‌کننده اثر می‌گذارند. در این مقاله یک دسته میله‌ی سوختی شامل 60 میله‌ی سوخت به همراه 4 شبکهی نگه‌دارنده مطابق با ابعاد اصلی رآکتور VVER-440 مدل‌سازی و به روش عددی حل شده است. از 3 مدل تلاطم استفاده شد. نتایج نشان داد در حالتی که از شبکه‌های نگه‌دارنده در طول دسته میله‌ی سوخت استفاده شود؛ ضریب انتقال حرارت در طول کانال نسبت به حالت بدون شبکهی نگه‌دارنده افزایش می‌یابد. البته با اضافه کردن این شبکه‌ها افت فشار نیز افزایش می‌یابد. نتایج حاصل از این پژوهش برای طراحی و ساخت مجتمع‌های سوخت و به دست آوردن پارامترهای ترموهیدرولیکی مربوط به بهینه‌سازی انتقال حرارت میله‌ها کاربرد دارد.
تازه های تحقیق
- M. Asgari, M. R. Abdi, M. Talebi, Survey of increase Reynolds number in thermo hydraulic parameters of fluid flow around fuel bundle of VVER-440 reactor with three spacer grids, 17th Iranian Nuclear Conference, Iran Uranium Processing & Nuclear Fuel Manufacturing Co, Isfahan (2010).
2. M. Asgari, M. R. Abdi, M. Talebi, H. Ahmadikia, Thermal hydvaulic simalation of fluid flow and heat transfer around fuel bundle of PWR reactor and sarvey the effect of spacer grids, 19th International Medanic Conterence, 15-16 (2010).
3. K. Ikeda, M. Hoshi, Development of Mitsubishi high thermal performance grid, JSME International Journal, 45(3) (2002).
4. E. Baglietto, H. Ninokata, A turbulence model study for simulating flow inside tight lattice rod bundles, Nuclear Engineering and Design, 235 (2005) 773–784.
5. D. Chang, S. Tavoularis, Simulations of turbulence, heat transfer and mixing across narrow gaps between rod-bundle subchannels, Nuclear Engineering and Design, 238 (2008) 109–123.
6. F. Baratto, S. Bailey, S. Tavoularis, Measurements of frequencies and spatial correlations of coherent structures in rod bundle flows, Nuclear Engineering and Design, 236 (2006) 1830–1837.
7. A. Aszodi, S. Toth, “CFD study on coolant mixing in VVER-440 fuel rod bundles and fuel assembly heads, Nuclear Engineering and Design, 240 (2009) 2194–2205.
8. A. Aszódi, S. Tóth, CFD analysis of flow field in a triangular rod bundle, Nuclear Engineering and Design, 21 (2008) 352–363.
9. C. Tzanos, Performance of k−ε turbulence models in the simulation of LWR fuel-bundle flows, Nuclear Engineering and Design, 84 (2001) 197–199.
10.S. Chang, A. Moon, Phenomenological investigations on the turbulent flow structures in a rod bundle array with mixing devices, Nuclear Engineering and Design, 238 (2006) 600–609.
11.Ansys, Fluent 6.3 user’s guide, (2006).
12.M. M. EL-Wakil, Nuclear energy conversion, 4th Edition, American Nuclear Society (1982).
13.Incropera, F. De Witt, Introduction to heat transfer, 4th Edition, USA (2002).
14.S. Tóth, A. Aszódi, Calculations of Coolant Flow in a VVER-440 Fuel Bundle with the Code Ansys CFX 10.0, Proc. Technical Meeting on Use of CFD Codes for Safety Assessment of Reactor Systems, Pisa Italy.
کلیدواژه‌ها
عنوان مقاله [English]
Numerical Simulation of Pressure Loss and Heat Transfer in Road Bundle Fuel Assembly with Spacer Grids
نویسندگان [English]
- M Asgari 1
- M Talebi 2
- M. R Abdi 3
چکیده [English]
The spacer grids in nuclear fuel assembly are one of the equipments that affect the fluid flow distribution with the creation of turbulence and driven fluid flow in a sub-channel. In the present paper, the fuel bundle in the VVER-440 nuclear reactor that contains 60 fuel rods and 4 spacer grids has been simulated and solved by a numerical method. Three turbulent models were used. The results showed that using spacer grids over the fuel bundle led to an increase in the heat transfer coefficients. However, these grids increase the pressure drop. The results of this research can be used to design and manufacture the fuel assembly and obtain the relevant thermo hydraulic parameters to optimize the heat transfer of fuel rods.
کلیدواژه‌ها [English]
- Spacer Grid
- Fuel Assembly
- Thermal-Hydraulic Parameters
- M. Asgari, M. R. Abdi, M. Talebi, Survey of increase Reynolds number in thermo hydraulic parameters of fluid flow around fuel bundle of VVER-440 reactor with three spacer grids, 17th Iranian Nuclear Conference, Iran Uranium Processing & Nuclear Fuel Manufacturing Co, Isfahan (2010).
2. M. Asgari, M. R. Abdi, M. Talebi, H. Ahmadikia, Thermal hydvaulic simalation of fluid flow and heat transfer around fuel bundle of PWR reactor and sarvey the effect of spacer grids, 19th International Medanic Conterence, 15-16 (2010).
3. K. Ikeda, M. Hoshi, Development of Mitsubishi high thermal performance grid, JSME International Journal, 45(3) (2002).
4. E. Baglietto, H. Ninokata, A turbulence model study for simulating flow inside tight lattice rod bundles, Nuclear Engineering and Design, 235 (2005) 773–784.
5. D. Chang, S. Tavoularis, Simulations of turbulence, heat transfer and mixing across narrow gaps between rod-bundle subchannels, Nuclear Engineering and Design, 238 (2008) 109–123.
6. F. Baratto, S. Bailey, S. Tavoularis, Measurements of frequencies and spatial correlations of coherent structures in rod bundle flows, Nuclear Engineering and Design, 236 (2006) 1830–1837.
7. A. Aszodi, S. Toth, “CFD study on coolant mixing in VVER-440 fuel rod bundles and fuel assembly heads, Nuclear Engineering and Design, 240 (2009) 2194–2205.
8. A. Aszódi, S. Tóth, CFD analysis of flow field in a triangular rod bundle, Nuclear Engineering and Design, 21 (2008) 352–363.
9. C. Tzanos, Performance of k−ε turbulence models in the simulation of LWR fuel-bundle flows, Nuclear Engineering and Design, 84 (2001) 197–199.
10.S. Chang, A. Moon, Phenomenological investigations on the turbulent flow structures in a rod bundle array with mixing devices, Nuclear Engineering and Design, 238 (2006) 600–609.
11.Ansys, Fluent 6.3 user’s guide, (2006).
12.M. M. EL-Wakil, Nuclear energy conversion, 4th Edition, American Nuclear Society (1982).
13.Incropera, F. De Witt, Introduction to heat transfer, 4th Edition, USA (2002).
14.S. Tóth, A. Aszódi, Calculations of Coolant Flow in a VVER-440 Fuel Bundle with the Code Ansys CFX 10.0, Proc. Technical Meeting on Use of CFD Codes for Safety Assessment of Reactor Systems, Pisa Italy.
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