نتائج البحث

An Efficient and Robust ILU(k) Preconditioner for Steady-State Neutron Diffusion Problem Based on MOOSE.

• Authors : Wu, Yingjie¹ (AUTHOR) ; Zhang, Han¹ (AUTHOR) ; Liu, Lixun¹ (AUTHOR)

• Source: Energies (19961073). Mar2024, Vol. 17 Issue 6, p1499. 16p.

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PARALLEL MESHLESS RADIAL BASIS FUNCTION COLLOCATION METHOD FOR NEUTRON DIFFUSION PROBLEMS.

• Authors : TANBAY, Tayfun¹

• Source: Uludag University Journal of the Faculty of Engineering (UUJFE). 2024, Vol. 29 Issue 1, p173-189. 17p.

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Residual resampling-based physics-informed neural network for neutron diffusion equations

• Authors : Zhang, Heng; He, Yun-Ling; Liu, Dong

Subjects: Computer Science - Machine Learning

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PARALLEL MESHLESS RADIAL BASIS FUNCTION COLLOCATION METHOD FOR NEUTRON DIFFUSION PROBLEMS

• Authors : Tayfun Tanbay

Subjects: paralel; ağsız; radyal baz fonksiyonu

• Source: Uludağ University Journal of The Faculty of Engineering, Vol 29, Iss 1, Pp 173-190 (2024)

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Apply implicitly restarted Arnoldi method to solving eigenvalue problem and reducing dimensionality in neutron diffusion

• Authors : XIANG Zhaocai; CHEN Qiafeng; ZHAO Pengcheng

Subjects: neutron diffusion equation; implicitly restarted arnoldi method; proper orthogonal decomposition

• Source: He jishu, Vol 47, Iss 2, Pp 135-141 (2024)

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Using Laplace Residual Power Series Method in Solving Coupled Fractional Neutron Diffusion Equations with Delayed Neutrons System.

• Authors : Shqair, Mohammed¹ (AUTHOR) ; Ghabar, Ibrahim¹ (AUTHOR); Burqan, Aliaa¹ (AUTHOR)

• Source: Fractal & Fractional. Mar2023, Vol. 7 Issue 3, p219. 11p.

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A posteriori error estimates for mixed finite element discretizations of the Neutron Diffusion equations.

• Authors : Ciarlet, Patrick¹; Do, Minh Hieu²; Madiot, François²

• Source: ESAIM: Mathematical Modelling & Numerical Analysis (ESAIM: M2AN). Jan/Feb2023, Vol. 57 Issue 1, p1-27. 27p.

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Reduced basis method for non-symmetric eigenvalue problems: application to the multigroup neutron diffusion equations

• Authors : Taumhas, Yonah Conjungo; Dusson, Geneviève; Ehrlacher, Virginie

Subjects: Mathematics - Numerical Analysis

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Uncertainty Analysis of Neutron Diffusion Eigenvalue Problem Based on Reduced-order Model

• Authors : In order to improve the efficiency of core physical uncertainty analysis based on sampling statistics, the proper orthogonal decomposition (POD) and Galerkin projection method were combined to study the application feasibility of reduced-order model based on POD-Galerkin method in core physical uncertainty analysis. The two-dimensional two group TWIGL benchmark question was taken as the research object, the key variation characteristics of the core flux distribution were extracted under the finite perturbation of the group constants of each material region, and the full-order neutron diffusion problem was projected on the variation characteristics to establish a reduced-order neutron diffusion model. The reduced-order model was used to replace the full-order model to carry out the uncertainty analysis of the group constants of the material region. The results show that the bias of the mathematical expectation of keff calculated by reduced-order and full-order models is close to 1 pcm. In addition, compared with the calculation time required for uncertainty analysis of full-order model, the analysis time of reduced-order model (including the calculation time of the full-order model required for the construction of reduced-order model) is only 11.48%, which greatly improves the efficiency of uncertainty analysis. The biases of mathematical expectation of keff calculated by reduced-order and full-order models based on Latin hypercube sampling and simple random sampling are less than 8 pcm, and under the same sample size, the bias from the Latin hypercube sampling result is smaller. From the TWIGL benchmark test results, under the same sample size, Latin hypercube sampling method is more recommended for POD-Galerkin reduced-order model.

Subjects: reduced-order model; proper orthogonal decomposition; galerkin projection

• Source: Yuanzineng kexue jishu, Vol 57, Iss 8, Pp 1584-1591 (2023)

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Calculation of Thermal Neutron Diffusion Length in a Light Water Media using MCNP Code

• Authors : M. Arkani

Subjects: mcnp code; neutron diffusion length; neutron flux

• Source: مجله علوم و فنون هسته‌ای, Vol 44, Iss 3, Pp 165-170 (2023)

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