Effect of FLiBe thermal neutron scattering on reactivity of molten salt reactor

Liu, Yafen; Li, Wenjiang; Zou, Yang; Yu, Shihe; Zhou, Bo; Cai, X.

doi:10.1051/epjconf/202023914008

Cited by 4 publications

(1 citation statement)

References 7 publications

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“…Scattering of thermal neutrons by moderators has a significant impact on reactor core design. It has been reported that thermal neutron scattering law (TSL) data of graphite have a large sensitivity to effective neutron multiplication factors (keff) of not only HTGRs [1] but also MSRs [2].…”

Section: Introductionmentioning

confidence: 99%

Evaluation of thermal neutron scattering law of nuclear-grade isotropic graphite

Nakayama,

Iwamoto,

Kimura

2024

EPJ Web of Conf.

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Graphite is a candidate of moderator in innovative nuclear reactors such as high-temperature gas-cooled reactors and molten salt reactors. Scattering of thermal neutrons by a moderator material has a significant impact on the reactor core design. To contribute to the development of the innovative nuclear reactors, thermal neutron scattering law data for nuclear-grade graphite were evaluated. Inelastic scattering component due to lattice vibration was evaluated based on phonon density of states computed with first-principles lattice dynamics simulations. The simulations were performed for ideal crystalline graphite. The evaluated inelastic scattering component well reproduced the experimental double-differential cross sections of scattered neutrons from nuclear-grade isotropic graphite measured at Materials and Life science experimental Facility (MLF) of J-PARC. Coherent elastic scattering component due to crystal structure was evaluated based on the neutron scattering and transmission experiments performed at MLF of J-PARC. Regarding comparison with the neutron transmission experiment, it was found that the quantification of small-angle neutron scattering due to structures larger than crystal, such as pores in graphite, is important.

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Section: Introductionmentioning

confidence: 99%

Evaluation of thermal neutron scattering law of nuclear-grade isotropic graphite

Nakayama,

Iwamoto,

Kimura

2024

EPJ Web of Conf.

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Assembly-level analysis on temperature coefficient of reactivity in a graphite-moderated fuel salt reactor fueled with low-enriched uranium

Cui

Zou

et al. 2023

NUCL SCI TECH

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To provide a reliable and comprehensive data reference for core geometry design of graphite-moderated and low-enriched uranium fueled molten salt reactors, the influences of geometric parameters on the temperature coefficient of reactivity (TCR) at an assembly level were characterized. A four-factor formula was introduced to explain how different reactivity coefficients behave in terms of the fuel salt volume fraction and assembly size. The results show that the fuel salt temperature coefficient (FSTC) is always negative owing to a more negative fuel salt density coefficient in the over-moderated region or a more negative Doppler coefficient in the under-moderated region. Depending on the fuel salt channel spacing, the graphite moderator temperature coefficient (MTC) can be negative or positive. Furthermore, an assembly with a smaller fuel salt channel spacing is more likely to exhibit a negative MTC. As the fuel salt volume fraction increases, the negative FSTC first weakens and then increases, owing to the fuel salt density effect gradually weakening from negative to positive feedback and then decreasing. Meanwhile, the MTC weakens as the thermal utilization coefficient caused by the graphite temperature effect deteriorates. Thus, the negative TCR first weakens and then strengthens, mainly because of the change in the fuel salt density coefficient. As the assembly size increases, the magnitude of the FSTC decreases monotonously owing to a monotonously weakened fuel salt Doppler coefficient, whereas the MTC changes from gradually weakened negative feedback to gradually enhanced positive feedback. Then, the negative TCR weakens. Therefore, to achieve a proper negative TCR, particularly a negative MTC, an assembly with a smaller fuel salt channel spacing in the under-moderated region is strongly recommended.

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