Design studies for A 50 MWth molten salt fast reactor
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Tarih
2023
Dergi Başlığı
Dergi ISSN
Cilt Başlığı
Yayıncı
Pergamon-Elsevier Science Ltd
Erişim Hakkı
info:eu-repo/semantics/closedAccess
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.
Açıklama
Anahtar Kelimeler
MSR, SMR, Fast reactor, Molten salt-fuel mixture, Liquid fuel
Kaynak
Progress in Nuclear Energy
WoS Q Değeri
Q1
Scopus Q Değeri
Q1
Cilt
166
