بررسی اثر مارپیچ خنک‌کننده رآکتور مینیاتوری بر افزایش زمان بهره‌برداری از رآکتور

زرگر, میثم; استکی, محمدحسین; طالبی, منصور; مختاری, جواد

doi:10.24200/nst.2023.1331

نوع مقاله : مقاله پژوهشی

نویسندگان

¹ گروه مهندسی هسته‌ای، دانشکده فیزیک، دانشگاه اصفهان، صندوق پستی: 3441-817467، اصفهان - ایران

² پژوهشکده رآکتور و ایمنی هسته‌ای، پژوهشگاه علوم و فنون هسته‌ای، صندوق پستی: 1589-81465، اصفهان - ایران

https://doi.org/10.24200/nst.2023.1331

چکیده

رآکتور مینیاتوری چشمه نوترونی یک رآکتور تحقیقاتی از نوع مخزن-استخری است که در آن از اورانیم با غنای بالابه‌عنوان سوخت، آب سبک به‌عنوان خنک‌کننده و برلیوم به‌عنوان بازتابنده استفاده ‌شده است. هنگام بهرهبرداری از رآکتور دمای سیال خنک‌کننده افزایش مییابد و باعث ایجاد فیدبک منفی در رآکتور می‌شود. بر اثر تزریق راکتیویته منفی در عین حال که یکی از مزایای این رآکتور و خصلت ایمنی ذاتی آن است، راکتیویته اضافی در دسترس کاهش‌یافته و زمان عملکرد رآکتور کاهش می‌یابد؛ بنابراین یکی از محدودیتهای اصلی رآکتور MNSR زمان عملکرد کوتاه آن (حدود 5_/2 ساعت) در قدرت اسمی است. در این پژوهش، یک تحلیل دینامیک سیالات محاسباتی برای مدل‌سازی پدیده جابهجایی آزاد در قلب، مخزن و استخر رآکتور انجام شده؛ و اثر کویل خنک‌کننده در بالای مخزن رآکتور بر افزایش زمان عملکرد رآکتور مورد بررسی قرار گرفته است. قلب رآکتور به‌عنوان یک محیط متخلخل و یک منبع حرارتی با توان ثابت 30 کیلووات در نظر گرفته شده است تا با کاهش جزییات، حجم محاسبات کاهش یابد. مقادیر تجربی و مقادیری که از حل عددی به دست آمده، تطابق قابل‌قبول دارند و این روش را تأیید می‌کنند. نتایج نشان می‌دهد؛ در حالت بدون کویل، افزایش دمای سیال خنک‌کننده به طور پیوسته یک شیب صعودی دارد؛ اما در حالت استفاده از کویل، افزایش دمای سیال خنک‌کننده تا مدت ‌زمان حدود 2 ساعت، شیب صعودی دارد اما بعد گذشت این زمان، شیب صعودی بسیار کم می‌شود و دما در یک محدوده نوسان می‌کند؛ در مقایسه با حالت بدون کویل، متوسط دما 3 درجه‌ کاهش‌یافته و زمان عملکرد رآکتور حدود 5_/1 ساعت افزایش پیدا می‌کند.

کلیدواژه‌ها

20.1001.1.17351871.1402.44.1.14.4

عنوان مقاله [English]

Investigating the effect of the miniature reactor cooling spiral on increasing the reactor operating time

نویسندگان [English]

M. Zargar ¹
M.H. Esteki ¹
M. Talebi ²
J. Mokhtari ²

¹ Department of Nuclear Engineering, Faculty of Physics, Isfahan University of Technology, P.O. Box: 817467-3441, Isfahan-Iran

² Reactor and Nuclear Safety Research School, Nuclear Science and Technology Research Institute, P.O.Box: 81465-1589, Isfahan - Iran

چکیده [English]

The Miniature Neutron Source Reactor is a tank-in-pool research reactor that uses highly-enriched uranium as fuel, light water as coolant, and beryllium as reflector. During the operation of the reactor, the coolant temperature increases and leads to negative feedback in the reactor. Although the negative reactivity is amongst the advantages of this reactor and an inherent safety feature; it causes a decrease in both the additional available reactivity and the reactor operation time. Therefore, short operation time (about 2.5 hours) at nominal power is one of the main limitations of the MNSR reactor. The natural convection in the core, vessel and pool of the reactor was modeled through a CFD analysis and the effect of the cooling coil at the top of the reactor tank on increasing the reactor operation time was investigated. To reduce the computations the details were decreased by considering the reactor core as a porous medium and a heat source with a constant power of 30kw. The experimental values and those obtained from the numerical solution are in good agreement. Results show a steady upward slope in the temperature rise of the coolant in the absence of coil, and an about-2-hours rise of the temperature in its presence. After this 2-hours period, the increasing rate decreases and the temperature fluctuates in a certain range. Compared to the case without the cooling coil, the average temperature is reduced by 3 degrees and the reactor operation time is increased by 1.5 hours.

کلیدواژه‌ها [English]

Cooling coil
Porous medium
CFD
Miniature neutron source reactor

مراجع

1. A. Hainoun, S. Alissa, Full-scale modelling of the MNSR reactor to simulate normal operation, transients and reactivity insertion accidents under natural circulation conditions using the thermal hydraulic code ATHLET, Nuclear Engineering and Design., 235(1), 33-52 (2005).

2. Z.H.O.U Yongmao, Safe private nuclear tool the miniature neutron source reactor, Transactions of the American Nuclear Society, 56(1), 253-259 (1998).

3. F. Faghihi, S.M. Mirvakili, Burn up calculations for the Iranian miniature reactor: a reliable and safe research reactor, Nuclear Engineering and Design, 239(6), 1000-1009 (2009).

4. M. Ahmadi, A. Rabiee, A. Pirouzmand, Development of a 3D thermohydraulic-neutronic coupling model for accident analysis in research miniature neutron source reactor (MNSR), Nuclear Engineering and Technology, 51(7), 1776-1783 (2019).

5. J. Al Zain, et al, Validation of DRAGON4/DONJON4 simulation methodology for a typical MNSR by calculating reactivity feedback coefficient and neutron flux, Results in Physics, 9, 1155-1160 (2018).

6. M. Albarhoum, S.A. Mohammed, Thermal-hydraulic code (THYD) for the miniature neutron source reactor thermal-hydraulic transients, Progress in Nuclear Energy, 51(3), 470-473 (2009).

7. I. Khamis, W. Alhalabi, Assessment of cooling effects on extending the maximum operating time for the Syrian miniature neutron source reactor, Progress in Nuclear Energy, 49(3), 253-261 (2007).

8. Y.A. Ahmed, et al, Effects of core excess reactivity and coolant average temperature on maximum operable time of NIRR-1 miniature neutron source reactor, Nuclear Engineering and Design, 241(5), 1559-1564 (2011).

9. M. Nazari, N. Babazadeh Baie, M.M. Shahmardan, Comparing Between Thermal Performance of Coiled Tube and Straight Tube Inserted in a Cylindrical Reservoir in the Transient Convection Heat Transfer, Experimental Study, Amirkabir Journal of Mechanical Engineering, 49(3), 557-566 (2017).

10. H. Mirgolbabaei, et al, Numerical estimation of mixed convection heat transfer in vertical helically coiled tube heat exchangers, International Journal for Numerical Methods in Fluids, 66(7), 805-819 (2015).

11. R. Chen, et al, Three dimensional thermal hydraulic characteristic analysis of reactor core based on porous media method, Annals of Nuclear Energ, 104, 178 (2017).

12. S. Sipaun, Shoaib Usman, Prediction of Missouri S&T Reactor's natural convection with porous media approximation, Nuclear Engineering and Design, 285, 241 (2015).

13. Y. Abbassi, et al, Investigation of natural convection in Miniature Neutron Source Reactor of Isfahan by applying the porous media approach, Nuclear Engineering and Design, 309, 213 (2016).

14. Y. Yizhou, Ph.D. Thesis, Illinois at Urbana-Champaign University, (2011). Retrieved from http://hdl.handle.net/2142/29660.

مجله علوم، مهندسی و فناوری هسته‌ای

بررسی اثر مارپیچ خنک‌کننده رآکتور مینیاتوری بر افزایش زمان بهره‌برداری از رآکتور

مراجع

مراجع

1. A. Hainoun, S. Alissa, Full-scale modelling of the MNSR reactor to simulate normal operation, transients and reactivity insertion accidents under natural circulation conditions using the thermal hydraulic code ATHLET, Nuclear Engineering and Design., 235(1), 33-52 (2005).

2. Z.H.O.U Yongmao, Safe private nuclear tool the miniature neutron source reactor, Transactions of the American Nuclear Society, 56(1), 253-259 (1998).

3. F. Faghihi, S.M. Mirvakili, Burn up calculations for the Iranian miniature reactor: a reliable and safe research reactor, Nuclear Engineering and Design, 239(6), 1000-1009 (2009).

4. M. Ahmadi, A. Rabiee, A. Pirouzmand, Development of a 3D thermohydraulic-neutronic coupling model for accident analysis in research miniature neutron source reactor (MNSR), Nuclear Engineering and Technology, 51(7), 1776-1783 (2019).

5. J. Al Zain, et al, Validation of DRAGON4/DONJON4 simulation methodology for a typical MNSR by calculating reactivity feedback coefficient and neutron flux, Results in Physics, 9, 1155-1160 (2018).

6. M. Albarhoum, S.A. Mohammed, Thermal-hydraulic code (THYD) for the miniature neutron source reactor thermal-hydraulic transients, Progress in Nuclear Energy, 51(3), 470-473 (2009).

8. Y.A. Ahmed, et al, Effects of core excess reactivity and coolant average temperature on maximum operable time of NIRR-1 miniature neutron source reactor, Nuclear Engineering and Design, 241(5), 1559-1564 (2011).

9. M. Nazari, N. Babazadeh Baie, M.M. Shahmardan, Comparing Between Thermal Performance of Coiled Tube and Straight Tube Inserted in a Cylindrical Reservoir in the Transient Convection Heat Transfer, Experimental Study, Amirkabir Journal of Mechanical Engineering, 49(3), 557-566 (2017).

10. H. Mirgolbabaei, et al, Numerical estimation of mixed convection heat transfer in vertical helically coiled tube heat exchangers, International Journal for Numerical Methods in Fluids, 66(7), 805-819 (2015).

11. R. Chen, et al, Three dimensional thermal hydraulic characteristic analysis of reactor core based on porous media method, Annals of Nuclear Energ, 104, 178 (2017).

12. S. Sipaun, Shoaib Usman, Prediction of Missouri S&T Reactor's natural convection with porous media approximation, Nuclear Engineering and Design, 285, 241 (2015).

13. Y. Abbassi, et al, Investigation of natural convection in Miniature Neutron Source Reactor of Isfahan by applying the porous media approach, Nuclear Engineering and Design, 309, 213 (2016).

دوره 44، شماره 1 - شماره پیاپی 103
فروردین 1402
صفحه 111-120

بررسی اثر مارپیچ خنک‌کننده رآکتور مینیاتوری بر افزایش زمان بهره‌برداری از رآکتور

مراجع

مراجع

1. A. Hainoun, S. Alissa, Full-scale modelling of the MNSR reactor to simulate normal operation, transients and reactivity insertion accidents under natural circulation conditions using the thermal hydraulic code ATHLET, Nuclear Engineering and Design., 235(1), 33-52 (2005).

2. Z.H.O.U Yongmao, Safe private nuclear tool the miniature neutron source reactor, Transactions of the American Nuclear Society, 56(1), 253-259 (1998).

3. F. Faghihi, S.M. Mirvakili, Burn up calculations for the Iranian miniature reactor: a reliable and safe research reactor, Nuclear Engineering and Design, 239(6), 1000-1009 (2009).

4. M. Ahmadi, A. Rabiee, A. Pirouzmand, Development of a 3D thermohydraulic-neutronic coupling model for accident analysis in research miniature neutron source reactor (MNSR), Nuclear Engineering and Technology, 51(7), 1776-1783 (2019).

5. J. Al Zain, et al, Validation of DRAGON4/DONJON4 simulation methodology for a typical MNSR by calculating reactivity feedback coefficient and neutron flux, Results in Physics, 9, 1155-1160 (2018).

6. M. Albarhoum, S.A. Mohammed, Thermal-hydraulic code (THYD) for the miniature neutron source reactor thermal-hydraulic transients, Progress in Nuclear Energy, 51(3), 470-473 (2009).

8. Y.A. Ahmed, et al, Effects of core excess reactivity and coolant average temperature on maximum operable time of NIRR-1 miniature neutron source reactor, Nuclear Engineering and Design, 241(5), 1559-1564 (2011).

9. M. Nazari, N. Babazadeh Baie, M.M. Shahmardan, Comparing Between Thermal Performance of Coiled Tube and Straight Tube Inserted in a Cylindrical Reservoir in the Transient Convection Heat Transfer, Experimental Study, Amirkabir Journal of Mechanical Engineering, 49(3), 557-566 (2017).

10. H. Mirgolbabaei, et al, Numerical estimation of mixed convection heat transfer in vertical helically coiled tube heat exchangers, International Journal for Numerical Methods in Fluids, 66(7), 805-819 (2015).

11. R. Chen, et al, Three dimensional thermal hydraulic characteristic analysis of reactor core based on porous media method, Annals of Nuclear Energ, 104, 178 (2017).

12. S. Sipaun, Shoaib Usman, Prediction of Missouri S&T Reactor's natural convection with porous media approximation, Nuclear Engineering and Design, 285, 241 (2015).

13. Y. Abbassi, et al, Investigation of natural convection in Miniature Neutron Source Reactor of Isfahan by applying the porous media approach, Nuclear Engineering and Design, 309, 213 (2016).

دوره 44، شماره 1 - شماره پیاپی 103فروردین 1402صفحه 111-120

دوره 44، شماره 1 - شماره پیاپی 103
فروردین 1402
صفحه 111-120