Design studies for A 50 MWth molten salt fast reactor

Şahin S., Şahin H. M., Tunç G., Şahiner H.

PROGRESS IN NUCLEAR ENERGY, cilt.166, 2023 (SCI-Expanded) identifier identifier

  • Yayın Türü: Makale / Tam Makale
  • Cilt numarası: 166
  • Basım Tarihi: 2023
  • Doi Numarası: 10.1016/j.pnucene.2023.104964
  • Dergi Adı: PROGRESS IN NUCLEAR ENERGY
  • Derginin Tarandığı İndeksler: Science Citation Index Expanded (SCI-EXPANDED), Scopus, Academic Search Premier, Compendex, Environment Index, INSPEC, Pollution Abstracts, Civil Engineering Abstracts
  • Anahtar Kelimeler: Fast reactor, Liquid fuel, Molten salt-fuel mixture, MSR, SMR
  • Gazi Üniversitesi Adresli: Evet

Özet

In this paper, design studies for a 50 MWth Molten Salt Reactor (MSR) have been carried out for the eutectic point of the molten salt mixture. Neutronic calculations are performed with the 19.75% enriched uranium and 100% 7Li isotope contained. The MCNP6 nuclear code was used with the ENDF/B-VIII nuclear data library to determine geometry dimensions and criticality. The time-evolution of Pu and other heavy isotopes in the reactor are calculated with the interface code MCNPAS. Four models are investigated with different reactor vessel materials: Model 1: Ni alloy (NiCrW-Hastelloy steel), Model 2: Beryllium, Model 3: Graphite and Model 4: Silicon Carbide (SiC). For the respective models, time-dependent criticality calculations are performed with a startup criticality value of keff =1.0262, 1.0298, 1.0404, and 1.0223. The 235U consumed for the corresponding models over the 10 years of reactor operation are 257.9 kg, 258.8 kg, 259.4 kg and 257.9 kg, respectively. At the same time, the 238U consumptions are 400.6 kg, 380.0 kg, 379.8 kg, and 400.9 kg, respectively. The amount of the higher quality new fuel (239Pu) produced for 10 years is calculated as 129.10 kg, 127.41 kg, 122.95 kg and 128.66 kg, for the respected models.