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Yahui Wang - One of the best experts on this subject based on the ideXlab platform.

  • ntkFoam: An OpenFOAM based Neutron Transport kinetics solver for nuclear reactor simulation
    Computers & Mathematics with Applications, 2019
    Co-Authors: Yahui Wang, Junhe Yang
    Abstract:

    Abstract Due to the complexity of detailed nuclear reactor numerical simulation, especially the complicated geometry and multi-physics coupling properties of the advanced reactor, a nuclear reactor kinetics solver for nuclear reactor engineering and design needs to be developed. Based on the open source C++ software OpenFOAM, this work establishes a Neutron Transport kinetics solver, namely ntkFoam, for the nuclear reactor kinetics simulation, from the governing equations to the detailed implementations. The coupling between multi-group Neutron Transport equations and the delayed Neutron precursor balance equations are considered with using the finite volume method. By introducing the inverse power method to the OpenFOAM, the k-eigenvalue problems are calculated, and by coupling the Neutron Transport and delayed Neutron precursor calculations using the Euler implicit scheme, the transient kinetics problems are simulated. Numerical results show that the proposed ntkFoam can simulate the multi-group Neutron Transport kinetics problem accurately and flexibility, and both the regular and irregular mesh configurations can be adopted. This work can provide some new perspectives and foundations to nuclear reactor coupling calculations.

  • Neutron Transport solution of lattice boltzmann method and streaming based block structured adaptive mesh refinement
    Annals of Nuclear Energy, 2018
    Co-Authors: Yahui Wang, Ming Xie
    Abstract:

    Abstract Simulation of Neutron Transport problem is the kernel of nuclear reactor physics, whose application, however, is limited by the exorbitant computational cost and complex geometry structure. This paper presents a lattice Boltzmann method (LBM) for multi-group Neutron Transport process and proposes a streaming-based block-structured adaptive-mesh-refinement (SSAMR) technique. The Neutron lattice Boltzmann equation is deduced from the Neutron Transport equation and the macroscopic Neutron diffusion equation can be recovered from Neutron lattice Boltzmann equation via the Chapman-Enskog expansion, which makes the kinetic significance of lattice Boltzmann equation clearly. The significance of relaxation time for Neutron LBM is further discussed for the first time, and the factors affecting the Neutron relaxation process are studied deeply also. After establishing the Neutron LBM, the SSAMR technique is applied to efficiently utilizing the computational resources of proposed LBM. To simply achieve the data communication between different meshes and eliminate the discontinuity of scalar Neutron flux, a data exchange technique based on the streaming process of LBM is adopted. Simulation results show that the proposed LBM can be applied to solving Neutron Transport process in all dimensions, and the SSAMR technique can not only effectively reduce the computational cost, but also be easily implemented. This work may provide some new perspectives for solving the Neutron Transport process and a powerful thought for large and complex engineering calculation.

  • A POD reduced-order model for resolving the Neutron Transport problems of nuclear reactor
    Annals of Nuclear Energy, 1
    Co-Authors: Yue Sun, Junhe Yang, Yahui Wang
    Abstract:

    Abstract Due to the high-dimensional integro-differential properties of the Neutron Transport equation (NTE) and the large-scale property of the nuclear reactor system, the detailed Neutron Transport simulation can be very time-consuming, which urges researchers to develop a fast and accurate technique. To improve this condition, this work develops a set of reduced-order models for the NTE (NTEROM) by combining the proper orthogonal decomposition (POD) and the Galerkin projection techniques to efficiently solve the Neutron Transport problems. Results of some typical benchmarks show that the proposed POD based NTEROM can estimate the Neutron Transport behaviors with high fidelity and outstanding computing efficiency. Furthermore, the prediction function of the proposed NTEROMs can be extended to approximately calculate the Neutron Transport problems with the parameters outside of the snapshots’ parameter range. This work can provide new perspectives and ideas on the issue of improving the efficiency of numerical calculations in nuclear reactor engineering.

Edward J. Allen - One of the best experts on this subject based on the ideXlab platform.

  • Stochastic Neutron Transport equations for rod and plane geometries
    Annals of Nuclear Energy, 2000
    Co-Authors: Wyatt D. Sharp, Edward J. Allen
    Abstract:

    Abstract In the derivation of the Neutron Transport equation, it is assumed that Neutron populations are large enough so that fluctuations in the Neutron population due to random Neutron interactions can be ignored. This assumption is removed in this investigation. The result is a system of stochastic differential equations that models the random behavior of Neutron Transport. Rod and plane geometries are considered in the present investigation. Isotropic scattering is assumed. Numerical procedures are developed and tested for solving these systems. The results are compared with Monte-Carlo calculations which confirm the accuracy of these stochastic Neutron Transport equations.

  • A new approach to discrete-ordinates Neutron Transport in plane geometry
    Transport Theory and Statistical Physics, 1991
    Co-Authors: K. Ganguly, Edward J. Allen, H. D. Victory
    Abstract:

    Abstract The space-angle-energy dependent integro-differential equation of the Neutron Transport theory cannot be exactly solved analytically in its entire generality. Over the past three decades, there have been many research efforts to develop numerically efficient approximate methods to solve the Neutron Transport equation. In this paper, we will present a new approach to the Neutron Transport equation, which is suitable for numerical compiitations. To understand the limitations of the existing popular approximate methods, we first consider the exact solution of the one-speed Transport equation in plane geometry

Olga Martin - One of the best experts on this subject based on the ideXlab platform.

  • Analytical approaches for solving Neutron Transport problems
    2010
    Co-Authors: Nikos E. Mastorakis, Olga Martin
    Abstract:

    The exact solutions for stationary and non-stationary Neutron Transport equations corresponding to various source functions are presented. The semi-infinite medium is considered to have a specular-reflecting boundary with angular dependent externally-incident flux. Some adjustments were made to agree the Neutron Transport theory with astrophysical applications.

  • Plenary lecture 2: analytical approaches for solving Neutron Transport problems
    2010
    Co-Authors: Olga Martin
    Abstract:

    The Neutron flux is obtained as the solution of a Boltzmann Transport equation. In its various integro-differential forms, this equation can be used to solve problems from the following fields: nuclear physics, astrophysics, radiative Transport and Transport of particles in porous media. In the literature there are several approaches based on the numerical methods as: the least squares method, the finite element method, Monte Carlo method, Fourier transform, Laplace transform, spherical harmonics or PN method, Jacobi polynomials approximation, truncated series of Chebyshev polynomials. In this paper we replace our Neutron Transport problem with an equivalent problem of radiative transfer. The exact solutions for stationary and non-stationary Neutron Transport equations corresponding to various source functions are presented. Some adjustments were made to agree the Neutron Transport theory with astrophysical applications.

Zhaoli Guo - One of the best experts on this subject based on the ideXlab platform.

  • Discrete unified gas kinetic scheme for steady multiscale Neutron Transport
    Journal of Computational Physics, 1
    Co-Authors: Xiafeng Zhou, Zhaoli Guo
    Abstract:

    Abstract A discrete unified gas kinetic schemes (DUGKS) is developed to solve the steady multidimensional and multigroup Neutron Transport problems based on the Boltzmann equation. Compared with the traditional Neutron Transport methods, this steady DUGKS (SDUGKS) has the asymptotic preserving properties and can give good numerical predictions for multiscale Transport ranging from optically thin to optically thick regimes (diffusive limit). In the SDUGKS a delta-form-based discrete ordinate ( S N ) framework is adopted to solve the steady Boltzmann Transport equation. But unlike the direct interpolation in the traditional numerical methods for Neutron Transport, the interface angular fluxes are obtained by integrating the steady Boltzmann equation along the Neutron Transport direction. The traditional step schemes or other robust schemes can be adopted to discretize the increment to ensure numerical stability. Based on the above special treatments, the sweeping and iterative strategies of the new developed SDUGKS are completely the same as those of the traditional steady S N methods, and thus SDUGKS can be easily implemented in the existing S N Transport codes, only by adding the residuals into the source terms. Several numerical tests are performed and the numerical solutions show that the results of the present SDUGKS agree well with the reference solutions for problems with different values of optical thickness, which indicates the advantages of the developed SDUGKS over the conventional methods. Overall, the SDUGKS can serve as a potential numerical tool for multiscale Neutron Transport problems covering a wide range of optical thickness.

Liangzhi Cao - One of the best experts on this subject based on the ideXlab platform.