Design studies for A 50 MWth molten salt fast reactor

1Citations
Citations of this article
2Readers
Mendeley users who have this article in their library.
Get full text

Abstract

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.

Cite

CITATION STYLE

APA

Şahin, S., Şahin, H. M., Tunç, G., & Şahiner, H. (2023). Design studies for A 50 MWth molten salt fast reactor. Progress in Nuclear Energy, 166. https://doi.org/10.1016/j.pnucene.2023.104964

Register to see more suggestions

Mendeley helps you to discover research relevant for your work.

Already have an account?

Save time finding and organizing research with Mendeley

Sign up for free