MTR benchmark static calculations with MCNP5 code

Bousbia-Salah, Anis; Benkharfia, Hocine; Kriangchaiporn, Nateekoool; Petruzzi, A.; D'Auria, Francesco Saverio; Ghazi, N.

doi:10.1016/j.anucene.2007.09.016

Cited by 14 publications

(5 citation statements)

References 3 publications

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“…In addition, it was found that there was no boiling in obstructed channel except for the hot channel because of lateral heat conduction of adjacent channels [11]. Bousbia-Salah et al calculated the neutron flux and power distribution of 10 MW MTR using MCNP5 code, and it showed that the MCNP5 was reliable for the plate type fuel core simulation and the calculation results were in good agreement with previous study [12]. Xoubi et al investigated the impact of enrichment on neutron flux in the in-core facility of 10 MW MTR with OpenMC and found the importance of flux trap calculation while considering the conversion of reactor core from HEU to LEU [13].…”

Section: Introductionsupporting

confidence: 59%

Thermal Hydraulic and Neutronics Coupling Analysis for Plate Type Fuel in Nuclear Reactor Core

Wang

et al. 2020

Science and Technology of Nuclear Installations

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The thermal hydraulic and neutronics coupling analysis is an important part of the high-fidelity simulation for nuclear reactor core. In this paper, a thermal hydraulic and neutronics coupling method was proposed for the plate type fuel reactor core based on the Fluent and Monte Carlo code. The coupling interface module was developed using the User Defined Function (UDF) in Fluent. The three-dimensional thermal hydraulic model and reactor core physics model were established using Fluent and Monte Carlo code for a typical plate type fuel assembly, respectively. Then, the thermal hydraulic and neutronics coupling analysis was performed using the developed coupling code. The simulation results with coupling and noncoupling analysis methods were compared to demonstrate the feasibility of coupling code, and it shows that the accuracy of the proposed coupling method is higher than that of the traditional method. Finally, the fuel assembly blockage accident was studied based on the coupling code. Under the inlet 30% blocked conditions, the maximum coolant temperature would increase around 20°C, while the maximum fuel temperature rises about 30°C. The developed coupling method provides an effective way for the plate type fuel reactor core high-fidelity analysis.

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Section: Introductionsupporting

confidence: 59%

Thermal Hydraulic and Neutronics Coupling Analysis for Plate Type Fuel in Nuclear Reactor Core

Wang

et al. 2020

Science and Technology of Nuclear Installations

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“…Unit cell based on SFE, used in IAEA-TECDOC-233 (Winkler and Zeis, 1980) and elsewhere (Bousbia-Salah et al, 2008) was modeled in WIMS. This methodology of using unit cell to generate burnup dependent number densities has its roots in the methodology which is being used in lattice cell calculation codes (usually transport theory based) coupled with 3D whole core codes, based on diffusion or transport theory.…”

Section: Computational Methodologymentioning

confidence: 99%

“…Standard values of the reactor design parameters have been used as quoted in literature (Bousbia-Salah et al, 2008). Analyses are performed for HEU core i.e.…”

Section: Computational Frameworkmentioning

confidence: 99%

“…The IAEA benchmark Material Test Reactor (MTR) (Haack, 1992;Winkler and Zeis, 1980) was set for validating and verification of tools and methodology needed to perform necessary calculations to convert High Enriched Uranium (HEU) reactors to Low Enriched Uranium (LEU) reactors, using computational tools including MTR-PC (MTR-PC, 2001). With the availability of advanced computational resources and tools, such benchmarks are being revisited by the nuclear community to re-assess their results (Bousbia-Salah et al, 2008).…”

Section: Introductionmentioning

confidence: 99%

“…This methodology used U-235, . Similar methodology, with varying number of isotopes included in burned fuel, was used to perform MTR benchmark static calculations with MCNP5 code (Bousbia-Salah et al, 2008). This methodology used only U-235, U-236, U-238, Pu-239, Pu-240, Pu-241, Pu-242, Xe-135 and Sm-149 (9 isotopes in total) on basis of number densities provided in original IAEA benchmark specification (Winkler and Zeis, 1980) for Monte Carlo methodology based codes.…”

Section: Introductionmentioning

confidence: 99%

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