Investigation of Neutronic and Thermal Performance Using UGD and MOX Fuel in VVER-1000 Nuclear Power Reactor

Abstract

When examining the safety and design features of nuclear power reactors, its thermal performance in addition to neutronic characteristics is important. In this study, the neutronic and thermal performances of VVER-1000 reactor with two different fuel assembly arrangements were examined. Those fuel assemblies named as YD1 and YD2 are composed of 3.7% enriched LEU and 3.6% enriched LEU with 4% Gd2O3 uranium-gadolinium (UGD) and 2%, 3%, 4.2% Pu and 3.6% enriched LEU with 4% Gd2O3 (MOXGD), respectively. The effects of using UGD and MOXGD fuel assembly arrangements on criticality and isotope transformations according to burnup rate were investigated by means of MCNP5 and MONTEBURNS2.0 nuclear code, correspondingly the temperature and enthalpy changes of the coolant along the hot channel were examined with the help of the COBRA-IV PC thermal analysis code. The results obtained from this study were compared with similar studies in the literature and it was observed that the obtained findings were in accordance with the literature.

Keywords

• VVER-1000
• MCNP
• power density
• COBRA

UGD ve MOX Yakıtı Kullanılarak VVER-1000 Nükleer Reaktöründe Nötronik Ve Termal Performansın İncelenmesi

Abstract

Nükleer güç reaktörlerinin güvenlik ve tasarım özellikleri incelenirken nötronik karakteristiklerinin yanı sıra termal performansları da önemlidir. Bu çalışmada, iki farklı yakıt demeti düzenine sahip VVER-1000 reaktörünün nötronik ve termal performansları incelenmiştir. YD1 ve YD2 olarak isimlendirilen bu yakıt demeti düzenleri sırasıyla %3,7 zenginlikli LEU ve %3,6 zenginlikli LEU ile %4 Gd2O3Uranyum-Gadolinyum (UGD) bileşimi ve %2, %3, %4,2 Pu ve %3,6 zenginlikli LEU ile %4 Gd2O3 içeren (MOX) yakıt bileşiminden meydana gelmektedir. UGD ve MOXGD yakıt kullanımının kritiklik ve yanma sonunda yakıt bileşimi değişimleri üzerine etkileri MCNP5 ve MONTEBURNS2.0 nükleer kodu yardımıyla incelenirken COBRA-IV PC termal analiz kodu yardımıyla sıcak kanal boyunca soğutucu akışkana ait sıcaklık ve entalpi değişimleri irdelenmiştir. Bu çalışmadan elde edilen sonuçlar literatürde yer alan benzer çalışmalarla karşılaştırılmış ve ulaşılan bulguların literatürle uyum içerisinde olduğu görülmüştür.

Keywords

• VVER-1000
• MCNP
• güç yoğunluğu
• COBRA

References

1. 1. Khan, S. A., Jagannathan, V., Kannan, U., Mathur, A.,” Study of VVER-1000 OECD LEU and MOX Computational Benchmark with VISWAM Code System”, Nuclear Energy and Technology 2, 312–334, (2016).
2. 2. Lazarenko, A., Kalugin, M., Bychkov, S., “Benchmark Calculations for VVER-1000 Fuel Assemblies Using Uranium or MOX Fuel” Institute of Energy and Nuclear Research,www.ipen.br/biblioteca/cd/physor/2000/physor/144.pdf
3. 3. Thilagam, L., Sunny, S. C., Jagannathan, V., Subbaiah, K. V., “A VVER-1000 LEU and MOX assembly computational benchmark analysis using the lattice burnup code EXCEL”, Annals of Nuclear Energy, 36(4): 505–519, (2009).
4. 4. Heba K. Louis, Esmat Amin,” The Effect of Burnup on Reactivity for VVER-1000 with MOXGD and UGD Fuel Assemblies Using MCNPX Code”, Journal of Nuclear and Particle Physics, 6(3): 61-71, (2016).
5. 5. Mercatali, L., Venturini. A., Daeubler, M., Sanchez, V. H., “SCALE and SERPENT solutions of the OECD VVER-1000 LEU and MOX burnup computational benchmark” Annals of Nuclear Energy, 83: 328–341, (2015).
6. 6. Dwiddar, M. S., Badawi, A. A., Abou-Gabal, H. H., El-Osery, I. A.,” Investigation of different scenarios of thorium–uranium fuel distribution in the VVER-1200 first core”, Annals of Nuclear Energy, 85: 605–612, (2015).
7. 7. Galahom, A. A.,” Reducing the plutonium stockpile around the world using a new design of VVER-1200 assembly”, Annals of Nuclear Energy, 119: 279–286, (2018).
8. 8. Galahom, A. A.,” Minimization of the fission product waste by using thorium based fuel instead of uranium dioxide”, Nuclear Engineering and Design, 314: 165–172, (2017).

9. 9. Genç, Y., Uzun, S., Acır, A. “VVER-1000 Nükleer Güç Reaktöründe Kritiklik ve Bağıl Güç Yoğunluk Dağılımının İncelenmesi”. Politeknik Dergisi, 23(4): 1379-1385, (2020). 10. Faghihi, F., Mirvakili, S.M., Safaei, S., Bagheri, S.,” Neutronics and sub-channel thermal-hydraulics analysis of the Iranian VVER-1000 fuel bundle”, Progress in Nuclear Energy 87, 39-46, (2016).
10. 11. Saadati, H., Hadad, K., Rabiee, A., “Safety margin and fuel cycle period enhancements of VVER-1000 nuclear reactor using water/silver nanofluid”, Nuclear Engineering and Technology, 50, 639-647, (2018).
11. 12. Nazari, T., Rabiee, A, Kazeminejad H., “Numerical investigation of the modal characteristics for a VVER-1000 fuel assembly”, Nuclear Engineering and Design, 345, 1-6, (2019).
12. 13. Nasr M. A., Zangian, M., Abbasi, M., Zolfaghari, A., “Neutronic and thermal-hydraulic aspects of loading pattern optimization during the first cycle of VVER-1000 reactor using Polar Bear Optimization method”, Annals of Nuclear Energy, 33, 538-548, (2019).
13. 14. Arshi, S. S., Mirvakili S. M., Faghihi, F. “Modified COBRA-EN Code to Investigate Thermal-Hydraulic Analysis of the Iranian VVER-1000 Core.” Progress in Nuclear Energy 52(6): 589–595, (2010).
14. 15. Trellue, H.R., 2003. Monteburns 2.0, An Automated, Multi-Step Monte CarloBurnup Code System, User’s Manual Version 2.0, Oak Ridge National Laboratory, PSR-455, (2003).
15. 16. Briesmeister J.F., A General Monte Carlo N-Particle Transport Code, MCNP-A General Monte Carlo N-Particle Transport Code, Version 5, vol. II: User’s Guide, LA-CP-03-0245, Los Alamos National Laboratory, (2005).
16. 17. Ludwig, S., 2002. Revision to ORIGEN2 – Version 2.2, Computer Code, ORNL/TM-7175.
17. 18. COBRA-IV PC: A Personal Computer Version of Cobra-IV-I For Thermal-Hydraulic Analysis of Rod Bundle Nuclear Fuel Elements and Cores, U.S. Department of Energy under Contract DE-AC06-76RLO 1830, B. Webb January 1988
18. 19. Katona, T. J.,” Long-Term Operation of VVER Power Plants”, Nuclear Power – Deployment, Operation and Sustainability book, Intech, (2011).
19. 20. Dourougie, C., Emmett, M.B., Gehin, J. C., & Lillie, R. A. (Jul 1999). Analysis of Weapons-Grade MOX VVER-1000 Benchmarks with HELIOS and KENO (ORNL/TM--1999/78). United States
20. 21. OECD NEA, A VVER-1000 LEU and MOX Assembly Computational Benchmark: Specification and Results, (2002)
21. 22. Tabadar, Z., Aghajanpour, S., Jabbari, M., Khaleghi, M., M. Hashemi-Tilehnoee, “Thermal-hydraulic analysis of VVER-1000 residual heat removal system using RELAP5 code, an evaluation at the boundary of reactor repair mode, Alexandria Engineering Journal, 57(3): 1249-1259, (2018).