Plutonium and Minor Actinides Utilization in FUJI-U1 Molten Salt Reactor

Wulandari, Cici; Waris, Abdul; Permana, Sidik; Pramuditya, Syeilendra

doi:10.1088/1742-6596/1204/1/012132

Cited by 5 publications

(3 citation statements)

References 7 publications

Supporting

Mentioning

Contrasting

Order By: Relevance

“…The results show that the relationship between the pipe diameter in the natural circulation system and the velocity of the fluid is inversely proportional. Simulations related to neutronic analysis and the fuel cycle of MSR have been widely performed [17][18][19][20][21][22][23][24]. Related simulations for the safety analysis of modified freeze plug melting on MSR have also been performed using the semiimplicit moving particle method [25].…”

Section: Introductionmentioning

confidence: 99%

Numerical Simulation of the Effect of Molten Salt Channel Number on Temperature Distribution in the Molten Salt Core Reactor

Dwiyulianti,

Yonanda,

Zulfalina

et al. 2024

J. Phys.: Conf. Ser.

View full text Add to dashboard Cite

A critical aspect in the design of a nuclear reactor that needs to be considered is thermal analysis. This parameter relates to molten salt circulation, which serves as a fuel, and coolant in Molten Salt Reactor (MSR) core reactors. This study simulates the temperature distribution performance of the MSR core by considering the effect of the number of molten salt channels with inlet temperature variations. Knowing the temperature distribution will predict the characteristics of heat transfer in the form of natural convection for the survival of the reactor. The reactor used a Molten Salt Breeder Reactor (MSBR) design developed by Oak Ridge National Laboratory (ORNL). The Ansys software was used to simulate a 3D model under steady-state conditions. The geometry was divided into two regions: solid (graphite material) and fluid (molten salt). The thickness of the moderator and radius of the molten salt channel were 0.0508 and 0.0208 m, respectively. The moderator height was 1.98 m. The number of molten salt channels varied up to four. The results showed a temperature difference of 9 to 15 K when the channel increased from one to four at 839 K inlet temperature.

show abstract

Section: Introductionmentioning

confidence: 99%

Numerical Simulation of the Effect of Molten Salt Channel Number on Temperature Distribution in the Molten Salt Core Reactor

Dwiyulianti,

Yonanda,

Zulfalina

et al. 2024

J. Phys.: Conf. Ser.

View full text Add to dashboard Cite

show abstract

“…Recent researches on thermal breeder MSR have been performed in several designs, such as ORNL-designed MSBR [5][6][7][8][9], Japanese MSR-FUJI [10][11][12], and the more recent Chinese Thorium Molten Salt Reactor (TMSR) [13][14][15][16], to name a few. Many of the aforementioned researches were focused on the fuel cycle and transition into the thorium fuel cycle.…”

Section: Introduction mentioning

confidence: 99%

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

et al. 2021

View full text Add to dashboard Cite

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.

show abstract

“…One of the advanced nuclear reactors that implemented the natural circulation system to circulate its fuel as a passive safety design is the Molten Salt Reactor 14,15 . As reported in papers, 16‐19 the fuel of Molten Salt Reactor is molten salt from a mixture of LiF‐BeF 2 ‐ThF 4 ‐(PuMA)F 4 or LiF‐BeF 2 ‐ThF 4 ‐ 233 UF 4 . Where, Thorium fuel ( 232 Th) provides an advantage for safety factors in reactor operations related to the void condition 20‐23 …”

Section: Introductionmentioning

confidence: 99%

Experimental and simulation approach of the loop geometry effect on the natural circulation system of the advanced nuclear reactor

2020

Self Cite

View full text Add to dashboard Cite

Summary Lately, natural circulation has become an exciting topic because it has been proposed and applied in some advanced nuclear reactors as a passive safety system. The aim of the present study is to examine the effect of the loop geometry on the natural circulation based cooling system of a given power source. The experimental approach was achieved by the construction of two natural circulation loops with vertical heater and vertical cooler. Both circulation loops were built with a same vertical length of 100 cm and different horizontal lengths, each of 50 and 100 cm, respectively. The heater was realized with a stainless pipe wrapped by a Nichrome wire, whereas the cooling system was designed and built as a pipe in 30 cm long square tube. Thus, the cooling water will circulate around the pipe loop inside the rectangular cooling chamber. The temperature data acquisition was achieved via an Arduino‐based system, controlling four K‐type thermocouple sensors. The numerical simulation part in the present work, was carried out by the COMSOL Multiphysics software, using the dedicated heat transfer and fluid dynamic module, namely the computational fluid dynamic module. The experimental data were considered to parameterize the COMSOL input model. The experimental results showed a clear difference between two passive cooling loops according to the horizontal length. The doublings of the horizontal and vertical lengths of the loop have a direct effect on the fluid flow rate.

show abstract

Plutonium and Minor Actinides Utilization in FUJI-U1 Molten Salt Reactor

Cited by 5 publications

References 7 publications

Numerical Simulation of the Effect of Molten Salt Channel Number on Temperature Distribution in the Molten Salt Core Reactor

Numerical Simulation of the Effect of Molten Salt Channel Number on Temperature Distribution in the Molten Salt Core Reactor

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

Experimental and simulation approach of the loop geometry effect on the natural circulation system of the advanced nuclear reactor

Contact Info

Product

Resources

About