The neutronic characteristics of thermal molten salt reactor

Khakim, Azizul; Rhoma, F.; Waluyo, A.; Suharyana, S.

doi:10.1063/5.0058949

Cited by 7 publications

(3 citation statements)

References 5 publications

Supporting

Mentioning

Contrasting

Order By: Relevance

“…The contributions of fission reactions caused by thermal, intermediate, and fast neutrons are 93.2%, 6.39%, and 0.42%, respectively. This has also been confirmed by A. Khakim [14]. As the control rod absorbers are positioned above the core, they absorb the neutron flux in the axial direction.…”

Section: Neutron Energy Spectrum On Samplessupporting

confidence: 57%

Estimation of Neutron and Prompt Photon Dose Rate Distribution in TMSR-500 Using McNp6

et al. 2022

View full text Add to dashboard Cite

Thorium Molten Salt Reactor-500 (TMSR-500), one of the Generation IV nuclear reactors, is designed by Thorcon International, Pte. Ltd, which is projected to be built in Indonesia. The reactor core is radially surrounded by B4C shielding, but not the upper part. As the silo hall sits above the reactor core and is accessible by reactor personnel, the dose rate must be calculated in the area to ensure the workers receive an annual dose below the acceptable limit. The dose rate from neutrons and photons as the result of fission reactions are the only sources to be calculated in this research, without taking the source from fission products into account. This research aims to obtain the dose rate distribution of neutrons and prompt photons using Monte Carlo code MCNP6. The reactor was assumed to operate at a nominal thermal power of 557 MWth. Dose rate calculation was obtained from flux Tally F4 and converted into dose rate using Dose Energy Dose Function (DEDF) factor. Conversion factors of flux to the dose were based on ICRP-21 and ANSI/ANS-6.1.1 1977. The result of the calculations showed that the distribution of neutron and prompt photon fluxes does not reach the silo hall.

show abstract

Section: Neutron Energy Spectrum On Samplessupporting

confidence: 57%

Estimation of Neutron and Prompt Photon Dose Rate Distribution in TMSR-500 Using McNp6

et al. 2022

View full text Add to dashboard Cite

show abstract

“…MCNP is a well-established code for simulating the neutronic aspects of nuclear reactors and has been extensively used for various types of reactors [25][26][27][28][29][30]. Although MCNP may not be the most suitable simulation tool for MSR due to its decoupling from thermal-hydraulic calculations, it has nonetheless been used for simulating various MSR designs, such as MSBR [6,31,32], MSR-FUJI [9], TMSR-500 [33][34][35], and Integral Molten Salt Reactor (IMSR) [36], with good agreement to the reference. The original PCMSR design was also simulated with MCNP and found to be in good agreement with the Serpent-2 code [23].…”

Section: Methodsmentioning

confidence: 99%

Neutronic Design Modification of Passive Compact-Molten Salt Reactor

Dwijayanto,

Harto

2024

Atom Indo.

View full text Add to dashboard Cite

Passive compact molten salt reactor (PCMSR) is a design concept of a molten salt reactor (MSR) currently under development in Universitas Gadjah Mada, Indonesia. It is designed as a thermal breeder reactor using thorium fuel cycle. However, our previous study shows that the original PCMSR design was incorrectly modelled, primarily overestimating its thorium breeding capability. To improve PCMSR neutronic design, we modified the core configuration by the addition of radial fuel channel layers previously nonexistent in original PCMSR core design in various configurations. Neutronic parameters of modified PCMSR geometries in the beginning of life (BOL) were simulated using MCNP6.2 radiation transport code with ENDF/B-VII.0 library. All variations of fuel layer addition show improvement in both temperature coefficient of reactivity (TCR) and breeding ratio (BR), with TCR values became more negative and BR values are larger than unity, ensuring proper breeding capability. Configuration Inner Core-Outer Blanket (IC-OB) achieves the largest BR and lowest doubling time (DT), whilst its TCR is an improvement from the original design. Therefore, IC-OB fuel layer configuration can be applied to redesign the original PCMSR and used in various design optimization scenarios.

show abstract

“…MCNP is capable of modelling a complex geometry without simplification, allowing more accurate prediction of the reactor physics characteristics. Previously, MCNP was used to calculate neutronic parameters of MSBR [36,37], Integral Molten Salt Reactor (IMSR) [38], MSR-FUJI [39], TMSR-500 [40,41], and PCMSR [42]. Therefore, MCNP can be considered to be suitable to calculate the neutronic parameters of an MSR, despite the calculation was performed in a quasistatic condition.…”

Section: Calculation Methodsmentioning

confidence: 99%

Transmutation of Transuranic Elements as Solid Coating in Molten Salt Reactor Fuel Channel

Dwijayanto,

Miftasani,

Harto

2023

Tri Dasa Mega

View full text Add to dashboard Cite

The accumulation of spent nuclear fuel (SNF) is presently considered as a hindrance of the massive deployment of nuclear power plant, especially regarding the transuranic (TRU) elements. Eliminating TRU through transmutation is one of the most feasible alternative as a technical solution to solve the issue. This study explores the possibility of TRU transmutation using molten salt reactor (MSR) in a heterogeneous configuration, where a solid TRU is coated inside the fuel channel filled with liquid salt fuel. Such configuration is proposed to allow higher TRU loading into fluoride salt mixture without compromising the safety of the reactor. TRU coating was applied in consecutively outward radial fuel channel layers with coating thicknesses of 2.5 mm and 5 mm. Calculation was performed using MCNP6.2 radiation transport code and ENDF/B-VII.0 neutron cross section library. From the results, TRU coating with smaller thickness and positioned closer to the centre of the core exhibit higher transmutation efficiency due to exposure to higher neutron flux. Highest transmutation efficiency was achieved at 67.93% after 160 days of burnup. This shows a potential of achieving highly efficient TRU using heterogeneous configuration in MSR core.

show abstract

The neutronic characteristics of thermal molten salt reactor

Cited by 7 publications

References 5 publications

Estimation of Neutron and Prompt Photon Dose Rate Distribution in TMSR-500 Using McNp6

Estimation of Neutron and Prompt Photon Dose Rate Distribution in TMSR-500 Using McNp6

Neutronic Design Modification of Passive Compact-Molten Salt Reactor

Transmutation of Transuranic Elements as Solid Coating in Molten Salt Reactor Fuel Channel

Contact Info

Product

Resources

About