Investigations on a typical small modular PWR using coupled neutronic-thermal-mechanical evaluations to achieve long-life cycle-length

Mirian, Seyed Farshad; Ayoobian, Navid

doi:10.1016/j.pnucene.2019.103176

Cited by 2 publications

(3 citation statements)

References 24 publications

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“…To select the optimum solution, effects of various parameters on the cost function must be analyzed. In this research, the purpose of optimization is to select a pair in which heat removal from the reactor core occurs with the maximum convective heat transfer coefficient α and the lowest pumping power W. In this regard, based on computing parameters (α, W), cost function is introduced as follow [7, 8, 17]: (7) where and are the weighting coefficients that are introduced to normalize the values and , since the latter vary in different ranges and have different dimensions.…”

Section: Cost Functionmentioning

confidence: 99%

“…Generally, neutronic and thermohydraulic simulations of the reactor core, and thermal-mechanical modeling of the fuel rod should be done to design a conceptual reactor. Coupled neutronic-thermal-mechanical evaluations of a typical small modular PWR were done in a recent article [7]; and as a final result, 8 pairs of (pitch-fuel rod diameter) were introduced.…”

Section: Introductionmentioning

confidence: 99%

“…The characteristics and features of the PWR reactor core and fuel rod are as follows [7]: Also, top views of the reactor core, fuel assembly and fuel rod, simulated in MCNP code are shown in Fig. 1.…”

Section: Introductionmentioning

confidence: 99%

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Thermo-Hydraulic Investigations of a Typical Small Modular PWR Using Computational Fluid Dynamics Modelling

Mirian

Ayoobian

2021

Therm. Eng.

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Computational Fluid Dynamics (CFD) is widely used in different industrial applications. In this research, the application of CFD in the thermo-hydraulic evaluation for a typical small modular Pressurized Water Reactor (PWR) was studied using ANSYS Fluent software. First, reactor core with different (pitchfuel rod diameter) was simulated using MCNP code. Subsequently, axial neutron heat flux was calculated in the hottest rod. In the following, different fuel channels were simulated using ANSYS workbench and corresponding mass flow rate according to the fluid outlet temperature was computed using Fluent. Then, thermohydraulic parameters including pumping power, convective heat transfer coefficient and turbulent intensity were calculated. Artificial neural network (ANN) coupled with genetic algorithm (GA) was used for optimization; and pair pitch -fuel rod diameter (0.012 m, 0.0072 m) was selected as the optimum value. Also, Critical Heat Flux (CHF) was computed with CFD-simulation, and compared with Tong CHF correlation and Groeneveld look-up table. A good agreement was observed between results, but CHF obtained from CFD simulation was more conservative. According to the results, Minimum Departure from Nucleate Boiling Ratio (MDNBR) was obtained as 2.12, which was compatible with its typical value. Accordingly, it could be concluded that the optimum reactor core was in the safe mode in the steady state conditions.

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Section: Cost Functionmentioning

confidence: 99%

Section: Introductionmentioning

confidence: 99%