Preliminary study on weapon grade uranium utilization in molten salt reactor miniFUJI

Aji, Indarta Kuncoro; Waris, Abdul

doi:10.1063/1.4895861

Cited by 3 publications

(3 citation statements)

References 5 publications

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“…( ) (8) By assuming the right hand side of ( 7), with the fission operator F, the eigen equation can be formulated as:…”

Section: Monte Carlo Methodsmentioning

confidence: 99%

“…An approach to replace 233 U with 235 U in 232 Th-based fuels has been implemented in [8]. The results show that the operation of 50 MWt in a mini-FUJI reactor for 5 years can be carried out with a minimum enrichment of 95 % 235 U with a UF 4 composition of 1.96 %.…”

Section: Literature Review and Problem Statementmentioning

confidence: 99%

“…This value indicates that the corrosiveness of LiF-BeF 2 -UF 4 is relatively smaller. Based on the research [7][8][9][10][11], identification of the use of fuel based on 235 U+ 238 U in the LiF-BeF 2 -UF 4 fuel mixture in the MSR FUJI-12 reactor needs to be done.…”

Section: Literature Review and Problem Statementmentioning

confidence: 99%

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Neutronic analysis on molten salt reactor FUJI-12 using 235U as fissile material in LiF-BeF2-UF4 fuel

Mabruri

Syarifah

Aji

et al. 2022

EEJET

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Neutronic analysis on the Molten Salt Reactor FUJI-12 using the fissile material 235U in LiF-BeF2-UF4 has been carried out. The problem faced in the use of thorium-based fuel is that the amount of 233U is small and not available in nature. 233U was produced through the 232Th breeding at a cost of $46 million/kg. That is a very high price when compared to 235U enrichment, which is only $100/kg. The MSR FUJI-12 used in this study is a generation IV reactor with a mixture of liquid salt fuel LiF-BeF2-ThF4-UF4 and thorium-based fuel (232Th+233U). In this study, neutronic analysis was carried out by replacing thorium-based fuel with uranium-based fuel (235U+238U). Neutronic analysis was performed using the OpenMC 0.13.0 code, which is a Monte Carlo simulation-based neutron analysis code. The nuclear data library used for neutronic calculations is ENDF B-VII/1. The fuel is used in a LiF-BeF2-UF4 molten salt mixture with three eutectic compositions: fuel 1, fuel 2, and fuel 3. Each fuel composition is optimized by enriching 235U in UF4 by 3 % to 8 %. The optimization results show the stability of the reactor criticality value, which is the main parameter so that the reactor can operate for the specified time. The optimization results show that fuel 1 cannot reach its optimal state in each variation of 235U enrichment. Fuel 2 and fuel 3 can reach optimal conditions at a minimum enrichment of 8 % and 7 % 235U. The results of the analysis of the distribution of the neutron flux in the reactor core show the distribution of nuclear reactions that occur in the core. The distribution of flux values in fuel 1 shows that the fission chain reaction is not running perfectly. Fuel 2 and fuel 3 are more stable by maintaining maximum flux at the center of the reactor core.

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“…( ) (8) By assuming the right hand side of ( 7), with the fission operator F, the eigen equation can be formulated as:…”

Section: Monte Carlo Methodsmentioning

confidence: 99%

Section: Literature Review and Problem Statementmentioning

confidence: 99%

See 1 more Smart Citation

Neutronic analysis on molten salt reactor FUJI-12 using 235U as fissile material in LiF-BeF2-UF4 fuel

Mabruri

Syarifah

Aji

et al. 2022

EEJET

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Worldwide activities

Edwards

2017

Molten Salt Reactors and Thorium Energy

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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

Variastuti,

Irwanto

2024

J. Phys.: Conf. Ser.

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Nuclear data libraries playian important role in the accuracy of neutronic aspect calculations, which determine various factors in a nuclear reactor design. Likewise, deterministic or Monte Carlo methods significantly affect the calculation results. Thus, the combination of nuclear data library and calculation methods needs to be studied carefully. This research will conduct neutronic analysis for the Molten Salt Reactor (MSR) design, one of the Generation IV reactor types with advanced features for future energy resources, using the Monte Carlo-based program OpenMC and different nuclear data libraries. MSR will be designed for a 250 MWth power capacity, operable for five years. The fuel used will be LiF-BeF2 as a coolant and ThF4-UF4 as the fuel. The compositions for each fuel will be optimized to achieve critical conditions during five years of operation. This research aims to evaluate the use of differentinuclear data libraries on reactor criticality conditions, using the nuclear data libraries ENDF/VIII-B, ENDF/VII.0, JEFF 3.3, and JENDL 5.0. To perform the calculation and analysis, MSR chain depletion modification has to be done for different nuclear data libraries. Then, evaluation will be conducted on effective multiplicationifactor (k-eff), neutron flux spectrum, and cross-section. The results of this research show that the use of JENDL 5.0 provides more optimal critical conditions. This is due to the completeness of JENDL 5.0 nuclear data covering 795 nuclide data and higher fission cross-section values. While ENDF/VIII-B has 557 nuclide data evaluations compared to ENDF/VII.0’s 423 nuclide data, more stable k-eff values are shown for the use of ENDF nuclear data libraries. This research is expected to significantly contribute to meeting Indonesia’s energy needs by providing efficient solutions in MSR design in neutronic aspects using the most accurate nuclear data library.

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Preliminary study on weapon grade uranium utilization in molten salt reactor miniFUJI

Cited by 3 publications

References 5 publications

Neutronic analysis on molten salt reactor FUJI-12 using 235U as fissile material in LiF-BeF2-UF4 fuel

Neutronic analysis on molten salt reactor FUJI-12 using 235U as fissile material in LiF-BeF2-UF4 fuel

Worldwide activities

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

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