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

Yıl 2021, Cilt: 24 Sayı: 4, 1557 - 1565, 01.12.2021

Sinem Uzun , Yasin Genç , Adem Acır

https://doi.org/10.2339/politeknik.781689

Öz

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.

Anahtar Kelimeler

VVER-1000, MCNP, power density, COBRA

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

Öz

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.

Anahtar Kelimeler

VVER-1000, MCNP, güç yoğunluğu, COBRA

Kaynakça

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