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This article has been cited by the following article(s):
Comparison of Neutronic Analysis for 250 MWth Molten Salt Reactor using Monte Carlo OpenMC Code with Different Nuclear Data Libraries of JENDL 5.0, ENDF/B-VII.1, ENDF/B-VIII.0, and JEFF 3.3
Validation of OpenMC Code Criticality Value Calculation for GFR Reactor with UN-PuN Fuel
Fajri Prasetya, Ahmad Muzaki Mabruri, Iklimatul Karomah, Ratna Dewi Syarifah, Indarta Kuncoro Aji and Nuri Trianti Journal of Physics: Conference Series 2734(1) 012065 (2024) https://doi.org/10.1088/1742-6596/2734/1/012065
Effect of Different Temperature and Nuclear Data Libraries in Criticality Calculations of 300 MWt Molten Salt Reactor
WITHDRAWN: Comparison of uranium plutonium nitride (U Pu N) and thorium nitride (Th N) fuel for 500 MWth Gas Cooled Fast Reactor (GFR) longlife without refueling