Evaluating the JEFF 3.1, ENDF/B-VII.0, JENDL 3.3, and JENDL 4.0 nuclear data libraries for a small 100 MWe molten salt reactor with plutonium fuel

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

doi:10.1007/s41365-022-01141-8

Cited by 8 publications

(2 citation statements)

References 35 publications

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“…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.

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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.

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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.

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“…Radionuclides have been proven to be powerful tracers for soil erosion investigation [2,3,4]. Plutonium and 137 Cs (T 1/2 = 30.17 a) are artificial radionuclides mainly produced by artificial nuclear activities [5,6,7,8,9]. 239 Pu (T 1/2 = 2.411 × 10 4 a) and 240 Pu (T 1/2 = 6563 a) were the most important plutonium isotopes.…”

Section: Introductionmentioning

confidence: 99%

First application of plutonium in soil erosion research on terraces

et al. 2023

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The spatial distributions of 239+240Pu and 137Cs in soils from Longji Rice Terraces were investigated to evaluate soil erosion. The activity concentrations of 239+240Pu and 137Cs in the surface soils of the paddy fields were in the range of 0.089–0.734 and 1.80–7.88 mBq/g, respectively. The activities of 239+240Pu and 137Cs showed very similar distribution trends, first increasing and then decreasing with increasing elevation. The 240Pu/239Pu atom ratios in the surface soils ranged from 0.162 to 0.232. The activities of 239+240Pu and 137Cs in the soil cores tended to be uniformly distributed within the plowed layer and declined exponentially below this depth. The mean soil erosion rates of Longji Rice Terraces estimated by 239+240Pu and 137Cs tracer methods were 5.44 t/(ha·a) and 5.16 t/(ha·a), respectively, which demonstrated that plutonium can replace 137Cs as an ideal tracer for soil erosion research in the future. Landform features are the main factors affecting the distribution of plutonium and 137Cs as well as soil erosion in the Longji Rice Terraces.

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Nitrate salt phase transition study for molten salt loop working fluid consideration

Hariyanto,

Permana,

Waris

et al. 2023

Heat Mass Transfer

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Evaluating the JEFF 3.1, ENDF/B-VII.0, JENDL 3.3, and JENDL 4.0 nuclear data libraries for a small 100 MWe molten salt reactor with plutonium fuel

Cited by 8 publications

References 35 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

First application of plutonium in soil erosion research on terraces

Nitrate salt phase transition study for molten salt loop working fluid consideration

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