Fuel cycle analysis of molten salt reactors based on coupled neutronics and thermal-hydraulics calculations

Zhou, Shengcheng; Yang, Won Sik; Park, Tongkyu; Wu, Hongchun

doi:10.1016/j.anucene.2017.10.040

Cited by 15 publications

(9 citation statements)

References 16 publications

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“…The neutron reaction rates is calculated by the coupled neutronics and thermal‐hydraulics simulation solver which was built using the OpenFOAM finite volume toolkit, 11,12 and the depletion solver determines the evolution of the material composition by solving the Bateman equation. The depletion solver is an improved version of the depletion module of FAMOS 8 . The microscopic cross‐section library is generated by Monte Carlo code OpenMC 13 …”

Section: Theory and Methodsmentioning

confidence: 99%

“…In this content, the method proposed by Zhou 8 is adopted in this work. The method assumes that the concentration of nuclides is uniformly distributed in the core and external loop respectively and the inventory of the nuclides in the core and external loop is coupled through the flow continuity conditions.…”

Section: Theory and Methodsmentioning

confidence: 99%

“…In addition, only a few days were covered in the simulation because of the limitation of the numerical method adopted 7 . Zhou et al analyzed the impact of the thermal hydraulics feedback effect on the depletion and the build‐up behavior of the actinides and fission products (FPs) in the Molten Slat Breeder Reactor 8 …”

Section: Introductionmentioning

confidence: 99%

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Analysis of the thermal feedback effect and nuclides flow effect on the fuel‐cycle characteristics of Molten Salt Reactors

Cao

Yang

et al. 2020

Int J Energy Res

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Summary The main feature of Molten Salt Reactors (MSRs) is the use of fluid fuel which continuously flows through the primary circuit. The adoption of fluid fuel in the MSRs leads to strong coupling between the neutronics, thermal‐hydraulics, fuel‐cycle, and salt reprocessing. To accurately model the special phenomena caused by fluid fuel, a code system named TANSY has been developed. The code system has the capability to simulate the online refueling, the online reprocessing, the flow of delayed neutron precursors and the drift of nuclides accurately. To verify the fuel‐cycle analysis capability of the code, neutronics benchmark problem of the Molten Salt Fast Reactor was simulated. The numerical solutions calculated by TANSY agree well with the reference results. After verification, it is applied to analyze the thermal feedback effect and nuclides flow effect on the characteristics of the fast‐spectrum MSRs. From the numerical results, it has been concluded that both effects have a small influence less than 1% on the time evolution of the actinides. But it is also found that the thermal feedback effect has a relatively high impact about 8% on the feeding rate of the fissile nuclide because of its influence on the criticality and neutron spectrum. The nuclides flow effect has a relatively high impact about 5% on the time evolution of the nonsoluble fission products such as 135Xe in the core because of the high reprocessing rate of such elements.

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Section: Theory and Methodsmentioning

confidence: 99%

Section: Theory and Methodsmentioning

confidence: 99%

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Analysis of the thermal feedback effect and nuclides flow effect on the fuel‐cycle characteristics of Molten Salt Reactors

Cao

Yang

et al. 2020

Int J Energy Res

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“…(3.27), the first term represents the total importance of prompt neutrons produced by fission and the second term represents the total importance of delayed neutrons produced by precursor decay. Therefore, the effective delayed neutron fraction for a circulating fuel can be calculated as [17,18]…”

Section: Adjoint Flux and Kinetics Parametersmentioning

confidence: 99%

Multiphysics Coupling of PROTEUS-NODAL and SAM for Molten Salt Reactor Simulation

Jaradat

Yang

Park

et al. 2020

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Verification tests of the coupled system of PROTEUS-NODAL and SAM were performed using the steady state and transient problems derived from the MSFR benchmark problem. Since the radial crossflow is neglected in the current SAM model, the effect of this simplification was first examined by comparing the steady state results with those obtained by a manually coupled calculation of PROTEUS-NODAL and ANSYS CFX. The SAM calculation used four parallel axial channels and CFX performed the full 3D CFD calculation in the cylindrical geometry of the MSFR core. Due to the neglect of the radial velocity field in the SAM calculation, SAM underestimated the axial velocity at the core center slightly, and this resulted in a slightly top-skewed power distribution: 0.1% overestimation in the upper part and 0.2% underestimation in the lower part of the core. The UTOP, ULOF, and ULOHS accidents of the MSFR transient benchmark were analyzed by including the outer loop in the SAM model. The results were compared with the PSI solutions from a coupled PARCS and TRACE calculation and the TUDelft solutions obtained from a coupled neutron diffusion and CFD calculation. In general, the power and core-averaged fuel temperature solutions of the coupled PROTEUS-NODAL and SAM calculations agreed well with the other solutions.

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“…It is not the purpose of this work to discuss the extensive literature on this topic; interested readers may refer to the excellent review by Ault [4]. Most existing suggestions for the configuration of thorium reactors use molten salts as medium and graphite as a moderator [5,6]. Furthermore, fast reactors simulations have been suggested, either based on molten salts [7,8], lead cooled [9], or water.…”

Section: Introductionmentioning

confidence: 99%

A Model Study of a Homogeneous Light-Water Thorium Reactor

Canepa

2020

Physics

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This work presents a computational study of a 232 Th -based homogeneous light-water reactor. Thorium reactors have been proposed as an alternative to the uranium fuel cycle since they exploit both the availability of thorium and its ability to afford fissile uranium isotopes by a sequence of neutron captures. Besides 233 U , as a result of the neutron captures, a significant amount of 234 U (36.3%) and 6.46% of 235 U are formed in the reactor under study. More importantly, the proposed simulation points out the possibility of a continuous withdrawal of the uranium isotopes without compromising the criticality and the power output of the reactor. This withdrawal affords the fissile material for the startup of reactors other than the first one, which requires a one-time only limited amount of fissile material. The significant molar fraction of the 234 U (0.17) in the extracted fuel does not pose a limitation on weapon proliferation, as a consequence of its high fission cross section for high-energy neutrons.

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Fuel cycle analysis of molten salt reactors based on coupled neutronics and thermal-hydraulics calculations

Cited by 15 publications

References 16 publications

Analysis of the thermal feedback effect and nuclides flow effect on the fuel‐cycle characteristics of Molten Salt Reactors

Analysis of the thermal feedback effect and nuclides flow effect on the fuel‐cycle characteristics of Molten Salt Reactors

Multiphysics Coupling of PROTEUS-NODAL and SAM for Molten Salt Reactor Simulation

A Model Study of a Homogeneous Light-Water Thorium Reactor

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