Burnup optimization of once-through molten salt reactors using enriched uranium and thorium

Tan, Meng-Lu; Zhu, Guifeng; Zhang, Zhengde; Zou, Yang; Yu, Xiaohan; Yu, Chenggang; Dai, Ye; Yan, Rui

doi:10.1007/s41365-022-00995-2

Cited by 12 publications

(11 citation statements)

References 26 publications

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“…As the EALF gradually increased to the fast spectrum region, the breeder of uranium and plutonium showed a clear advantage, and the burnup starts to increase. On the whole, the burnup in the thermal spectrum region has a local maximum value in the range of VF = 10% − 15%, the local maximum is approximately 8 MW•d/kgU nat , similar to the finding of Tan et al [16]. The burnup has a local minimum value of burnup in the epithermal spectrum region, which is not conducive to the utilization of fuel.…”

Section: Resultssupporting

confidence: 83%

“…International Journal of Energy Research slowed state, and the graphite absorption accounts for a large proportion, which is not conducive to fuel utilization. When VF is large (about more than 20%), the neutron absorption reactivity of 238 U increases significantly caused by the hardening of the energy spectrum, which is not conducive to the fission absorption reaction, and the burnup begins to decrease [16]. As the EALF gradually increased to the fast spectrum region, the breeder of uranium and plutonium showed a clear advantage, and the burnup starts to increase.…”

Section: Resultsmentioning

confidence: 99%

“…This study adjusts the uranium enrichment of the initial fuel salt according to the set value of k inf or NLR, and the uranium enrichment of makeup salt is 20 wt%. According to the characteristics of molten salts, different molten salt temperatures are set, and the temperatures of the three molten salts are 900 K [16], 1023 K [18], and 960 K [19] in turn. Because of the limited solubility of FLiBe carrier salt for heavy metal nuclides, the upper limit of HN in the burnup process is set to 12% [15].…”

Section: Models and Methodsmentioning

confidence: 99%

“…Before this, fuel utilization studies for small modular molten salt reactors have been performed [15]. The fuel utilization of molten salt reactors is optimized and analyzed according to different molten salt components and volume fractions [16,17], but only in the case of thermal spectrum under the once-through fuel cycle mode. Furthermore, Small Modular Molten Salt Reactor design is one of the key directions in the development of molten salt reactors.…”

Section: Introductionmentioning

confidence: 99%

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Comparative Study on the Neutronic Performance of Thermal and Fast Molten Salt Reactors under Once-Through Fuel Cycle

Zhu

Liu

et al. 2023

International Journal of Energy Research

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Once-through molten salt reactors have the advantages of easy availability of fuel, nuclear nonproliferation, and low technical difficulty. It is expected to be the earliest commercial molten salt reactor fuel cycle mode. The current research on the neutronic performance of the once-through fuel cycle mainly focuses on the thermal spectrum range, and broader spectrum research needs to be carried out further. In particular, the fuel utilization characteristics of molten salt fast reactors in once-through fuel cycle are not yet clear. In addition, small modular design is one of the trends of molten salt reactor development, and small modular design under thermal and fast spectrum has different advantages and needs further comparative analysis. In this paper, the single lattice, bare reactor, and core with reflector models are used to compare the neutronic performance in once-through fuel cycle from thermal to fast spectrum, which includes burnup, variation of fuel salt composition and volume with burnup, temperature reactivity coefficient, neutron irradiation lifetime, and small modular size. The results show that the burnup increases first, then decreases, and then increases again with the increase of the average lethargy causing fission (EALF). For small-scale molten salt reactors, the difference in fuel utilization under thermal and fast spectra is small. However, for large-scale reactors, the burnup under fast spectrum is significantly higher than that under thermal spectrum. When the EALF is large enough, and the neutron loss ratio is small enough, the breed-and-burn (BNB) fuel cycle mode can be realized, and then, the fuel utilization will be significantly improved. In the fast spectrum reactors, the HN and volume of molten salt change violently because of the longer burnup period. Based on the infinite single lattice model, the temperature reactivity coefficient increases first, then decreases, and increases again with EALF. It is negative over the entire energy spectrum. In addition, the small modular designs are discussed by the bare reactor model, and the neutron irradiation lifetime of the materials is analyzed by the core model with a reflector.

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Section: Resultssupporting

confidence: 83%

Section: Resultsmentioning

confidence: 99%

Section: Models and Methodsmentioning

confidence: 99%

Section: Introductionmentioning

confidence: 99%

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Comparative Study on the Neutronic Performance of Thermal and Fast Molten Salt Reactors under Once-Through Fuel Cycle

Zhu

Liu

et al. 2023

International Journal of Energy Research

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“…[2][3][4] The ssile material (UF 4 ) and fertile material (ThF 4 ) are dissolved in the molten carrier salt, and several alternative fuel carrier salts are being considered, but the leading candidate is 2LiF-BeF 2 (FLiBe). [5][6][7][8] Using liquid fuel will give MSRs an excellent advantage in performing continuous on-line fuel purication. On-line refueling signicantly reduces the costs associated with fuel fabrication.…”

Section: Introductionmentioning

confidence: 99%

Effect of oxide impurities on the dissolution behavior of Th⁴⁺, Be²⁺ and U⁴⁺ in fluoride salts

Yan,

Li,

et al. 2024

RSC Adv.

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Influence of fiber type and amount of reinforcing fibers on radon exhalation in uranium tailings’ overburden under high temperature environment

Chen

Song

et al. 2023

J Radioanal Nucl Chem

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Burnup optimization of once-through molten salt reactors using enriched uranium and thorium

Cited by 12 publications

References 26 publications

Comparative Study on the Neutronic Performance of Thermal and Fast Molten Salt Reactors under Once-Through Fuel Cycle

Comparative Study on the Neutronic Performance of Thermal and Fast Molten Salt Reactors under Once-Through Fuel Cycle

Effect of oxide impurities on the dissolution behavior of Th⁴⁺, Be²⁺ and U⁴⁺ in fluoride salts

Influence of fiber type and amount of reinforcing fibers on radon exhalation in uranium tailings’ overburden under high temperature environment

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Burnup optimization of once-through molten salt reactors using enriched uranium and thorium

Cited by 12 publications

References 26 publications

Comparative Study on the Neutronic Performance of Thermal and Fast Molten Salt Reactors under Once-Through Fuel Cycle

Comparative Study on the Neutronic Performance of Thermal and Fast Molten Salt Reactors under Once-Through Fuel Cycle

Effect of oxide impurities on the dissolution behavior of Th4+, Be2+ and U4+ in fluoride salts

Influence of fiber type and amount of reinforcing fibers on radon exhalation in uranium tailings’ overburden under high temperature environment

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Product

Resources

About

Effect of oxide impurities on the dissolution behavior of Th⁴⁺, Be²⁺ and U⁴⁺ in fluoride salts