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

Creative Commons License

Uzun S., Genc Y., ACIR A.

JOURNAL OF POLYTECHNIC-POLITEKNIK DERGISI, cilt.24, sa.4, ss.1557-1565, 2021 (ESCI) identifier

  • Yayın Türü: Makale / Tam Makale
  • Cilt numarası: 24 Sayı: 4
  • Basım Tarihi: 2021
  • Doi Numarası: 10.2339/politeknik.781689
  • Dergi Adı: JOURNAL OF POLYTECHNIC-POLITEKNIK DERGISI
  • Derginin Tarandığı İndeksler: Emerging Sources Citation Index (ESCI), TR DİZİN (ULAKBİM)
  • Sayfa Sayıları: ss.1557-1565
  • Anahtar Kelimeler: VVER-1000, MCNP, power density, COBRA, HYDRAULIC ANALYSIS, CODE
  • Gazi Üniversitesi Adresli: Evet

Özet

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.