Coupling MCNP6/ANSYS codes to calculate axial temperature and power distribution in a VVER-1000 fuel assembly
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Riham M. Refeat
Abstract
The precise simulation of the reactor core either in steady state or during transients is very important for the safety assessment of nuclear power plants. This requires accurate determination of the parameters that influence the reactor operation. Coupling neutronic and thermal hydraulic schemes are developed to calculate these parameters. In the present paper, a coupling scheme between MCNP6 and ANSYS-FLUENT17.2 codes is proposed to obtain accurate radial and axial temperature distribution and hence pin power distribution for VVER-1000 fuel assembly. The Performance of the developed coupling scheme is investigated in steady state calculations. An iterative process is associated with the exchange of data between codes to meet the convergence criteria. The results obtained demonstrate that the proposed coupling scheme is able to simulate the VVER-1000 fuel assembly accurately. It gives information about thermal and neutronic behavior of the assembly and allows the feedback effects to be accurately modeled. This work is a step forward to establish a consistent methodology to be used in transient calculations.
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Author contributions: All the authors have accepted responsibility for the entire content of this submitted manuscript and approved submission.
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Research funding: None declared.
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Conflict of interest statement: The authors declare no conflicts of interest regarding this article.
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Artikel in diesem Heft
- Frontmatter
- A study of RCS depressurization strategy of CPR1000 SAMG
- Drones application scenarios in a nuclear or radiological emergency
- Safety analysis for integrity enhancement in nuclear power plants (NPPs) in case of seashore region site
- Frictional wear characteristics of nickel-based alloy and reactor material in pressure vessel reactor
- Improving FNMC for the matrix effect of spherical shell plutonium samples
- Calculation of core neutronic parameters in electron accelerator driven subcritical TRIGA reactor
- Preliminary study on TRU transmutation in VVER-1000 fuel assembly using MCNP6
- Research on the application of 22Na radiolocation detection technology in advanced manufacturing process control
- Coupling MCNP6/ANSYS codes to calculate axial temperature and power distribution in a VVER-1000 fuel assembly
- Experimental and theoretical investigation of forced convection heat transfer with CNTs and CuO water based nano-fluids
- Thermal hydraulic characteristics of silicon irradiation in a typical MTR reactor
- Photon dosimetry using selective data sampling with Particle Swarm optimization algorithm based on NaI(Tl) scintillation detector
- Calendar of events
Artikel in diesem Heft
- Frontmatter
- A study of RCS depressurization strategy of CPR1000 SAMG
- Drones application scenarios in a nuclear or radiological emergency
- Safety analysis for integrity enhancement in nuclear power plants (NPPs) in case of seashore region site
- Frictional wear characteristics of nickel-based alloy and reactor material in pressure vessel reactor
- Improving FNMC for the matrix effect of spherical shell plutonium samples
- Calculation of core neutronic parameters in electron accelerator driven subcritical TRIGA reactor
- Preliminary study on TRU transmutation in VVER-1000 fuel assembly using MCNP6
- Research on the application of 22Na radiolocation detection technology in advanced manufacturing process control
- Coupling MCNP6/ANSYS codes to calculate axial temperature and power distribution in a VVER-1000 fuel assembly
- Experimental and theoretical investigation of forced convection heat transfer with CNTs and CuO water based nano-fluids
- Thermal hydraulic characteristics of silicon irradiation in a typical MTR reactor
- Photon dosimetry using selective data sampling with Particle Swarm optimization algorithm based on NaI(Tl) scintillation detector
- Calendar of events
