Neutronics Assessment of Accident-Tolerant Fuel in Advanced Power Reactor 1400 (APR1400)

Hartanto, Donny; Alshamsi, Alya; Alsuwaidi, A.; Bilkhair, A.; Hukal, H.A.; Zubair, Muhammad

doi:10.17146/aij.2020.1073

Cited by 8 publications

(1 citation statement)

References 17 publications

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“…The second code was Serpent-2. It is a 3D continuous energy Monte Carlo neutron and transport code developed at VTT Technical Research Center of Finland, Ltd [24][25][26]. Serpent-2 has been employed previously to model SD-TMSR and MSBR, both showing good agreement with the previous results [7,27].…”

Section: Methodsmentioning

confidence: 85%

Model Comparison of Passive Compact-Molten Salt Reactor Neutronic Design Using MCNP6 and Serpent-2

et al. 2021

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Passive Compact Molten Salt Reactor (PCMSR) is a thermal breeder molten salt reactor (MSR) developed in Universitas Gadjah Mada, Indonesia, run in thorium fuel cycle. Its design was initially developed using deterministic code SRAC2006 but has never been compared with other codes. This paper attempts to compare PCMSR neutronic design using Monte Carlo codes MCNP6 and Serpent-2 with ENDF B/VII.0 continuous neutron cross-section library. The reactor was run in a pure thorium fuel cycle with lithium fluoride as its carrier salt. The analyzed parameters were effective multiplication factor (k eff ), temperature coefficient of reactivity (TCR), void coefficient of reactivity (VCR), and conversion ratio (CR). The result shows that there are several important discrepancies between the original calculation and this research. The Monte Carlo calculations implied that PCMSR core was able to be critical using lower fissile concentration than previously designed, but failed to reach CR above unity. While the TCR value was found to be negative, the VCR value was positive up until the 10 % void fraction. The PCMSR core suffered from ineffective neutron moderation and high neutron leakage. These findings imply that the previous PCMSR neutronic design is inaccurate. For PCMSR to be able to operate as a thermal breeder MSR, geometrical modifications must be performed to improve neutron moderation and reduce neutron leakage.

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