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Han Y.,Shanghai JiaoTong University | Jiang X.,Shanghai NuStar Nuclear Power Technology Co. | Wang D.,Shanghai JiaoTong University
Nuclear Engineering and Design | Year: 2014

Coarse Mesh Finite Difference (CMFD) has been widely adopted as an effective way to accelerate the source iteration of transport calculation. However in a core with hexagonal assemblies there are non-hexagonal meshes around the edges of assemblies, causing a problem for CMFD if the CMFD equations are still to be solved via tri-diagonal matrix inversion by simply scanning the whole core meshes in different directions. To solve this problem, we propose an unequal mesh CMFD formulation that combines the non-hexagonal cells on the boundary of neighboring assemblies into non-regular hexagonal cells. We also investigated the alternative hardware acceleration of using graphics processing units (GPU) with graphics card in a personal computer. The tool CUDA is employed, which is a parallel computing platform and programming model invented by the company NVIDIA for harnessing the power of GPU. To investigate and implement these two acceleration methods, a 2-D hexagonal core transport code using the method of characteristics (MOC) is developed. A hexagonal mini-core benchmark problem is established to confirm the accuracy of the MOC code and to assess the effectiveness of CMFD and GPU parallel acceleration. For this benchmark problem, the CMFD acceleration increases the speed 16 times while the GPU acceleration speeds it up 25 times. When used simultaneously, they provide a speed gain of 292 times. © 2014 Elsevier B.V. Source

Lu D.,Shanghai JiaoTong University | Yu L.,Shanghai JiaoTong University | Zhang S.,Shanghai NuStar Nuclear Power Technology Co.
Annals of Nuclear Energy | Year: 2015

In a conventional coarse mesh nodal method the more accurate treatment of intra-nodal axial heterogeneity requires iterative axial node re-homogenization using axial flux profiles either reconstructed from core-wise coarse mesh solution or obtained from channel-wise axial fine mesh calculation. In this paper a new nodal method formulation, using Channel-wise Intrinsic Axial Mesh Adaptation (CIAMA), is proposed to solve this problem in a more fundamental way. For a given transverse (radial) leakage, along each axial channel a rigorous sub-node heterogeneous calculation is performed with the explicit axial heterogeneity within each coarse axial node. However, the transverse leakage between the axial channels is still calculated on the basis of coarse axial nodes, using the axially averaged radial current in each coarse axial node. Since the coupling between the axial channels is through the coarse axial nodes, it is not necessary to match the boundaries of the axial sub-nodes of neighboring axial channels in order to incorporate the axial sub-node calculation as an intrinsic part of the whole core global calculation. Therefore in the CIAMA nodal method, each axial channel is allowed to have its own sub-nodes adapting to its own axial heterogeneity variation. The CIAMA method has been implemented in the commercial code EGRET, which is used to qualify CIAMA. Excellent results of modeling fuel grid and control rod movement are presented. Application of CIAMA to three-dimensional pin-by-pin core calculation is also discussed and demonstrated to work well. © 2015 Elsevier Ltd. All rights reserved. Source

Lv D.,Shanghai JiaoTong University | Zhang S.,Shanghai NuStar Nuclear Power Technology Co.
Hedongli Gongcheng/Nuclear Power Engineering | Year: 2014

The incapability to handle the heterogeneity introduced by spacer grid and partially-inserted control rod has long been considered as one of the noticeable defects for the conventional coarse-mesh nodal methods. To improve this defect, two new nodal methods with the capability to explicitly handle these heterogeneities in the framework of coarse-mesh nodal methods are proposed. Numerical results for IAEA 3D benchmark problem and the practical power reactor problems demonstrate that the proposed sub-mesh method is quite successful, it is able to effectively handle not only the partially inserted control rod, but also the fine spacer grid within an axial coarse mesh. ©, 2014, Yuan Zi Neng Chuban She. All right reserved. Source

Yu L.,Shanghai JiaoTong University | Lu D.,Shanghai JiaoTong University | Zhang S.,Shanghai NuStar Nuclear Power Technology Co. | Wang D.,Shanghai JiaoTong University
Annals of Nuclear Energy | Year: 2014

A group-decoupled direct fitting method is developed for multi-group pin power reconstruction, which avoids both the complication of obtaining 2D analytic multi-group flux solution and any group-coupled iteration. A unique feature of the method is that in addition to nodal volume and surface average fluxes and corner fluxes, transversely-integrated 1D nodal solution flux profiles are also used as the condition to determine the 2D intra-nodal flux distribution. For each energy group, a two-dimensional expansion with a nine-term polynomial and eight hyperbolic functions is used to perform a constrained least square fit to the 1D intra-nodal flux solution profiles. The constraints are on the conservation of nodal volume and surface average fluxes and corner fluxes. Instead of solving the constrained least square fit problem numerically, we solve it analytically by fully utilizing the symmetry property of the expansion functions. Each of the 17 unknown expansion coefficients is expressed in terms of nodal volume and surface average fluxes, corner fluxes and transversely-integrated flux values. To determine the unknown corner fluxes, a set of linear algebraic equations involving corner fluxes is established via using the current conservation condition on all corners. Moreover, an optional slowing down source improvement method is also developed to further enhance the accuracy of the reconstructed flux distribution if needed. Two test examples are shown with very good results. One is a four-group BWR mini-core problem with all control blades inserted and the other is the seven-group OECD NEA MOX benchmark, C5G7. © 2014 Elsevier Ltd. All rights reserved. Source

Peng S.,Shanghai JiaoTong University | Jiang X.,Shanghai JiaoTong University | Zhang S.,Shanghai JiaoTong University | Zhang S.,Shanghai NuStar Nuclear Power Technology Co. | Wang D.,Shanghai JiaoTong University
Annals of Nuclear Energy | Year: 2013

Subgroup parameters based on the intermediate resonance (IR) approximation are generated by the constrained best fitting method to assure the positivity of the subgroup parameters, and the infinite mass scattering term introduced by the IR approximation is explicitly retained in the subgroup equation. The impact on the heterogeneous problem is investigated. To preserve the reference reaction rates obtained by the subgroup calculation, the Hébert SPH method is adopted. A resonance interference correction factor table is introduced to consider the resonance interference effect. The proposed resonance calculation method is tested on various problems, which cover homogeneous cases, 1D pin cell cases, as well as 2D assembly cases. It is demonstrated that (1) the proposed resonance interference factor table method can significantly reduce the error caused by resonance interference; (2) the SPH method can reduce the error of subgroup collapsing; (3) the infinite mass scattering term can significantly impact the result of a heterogeneous problem. Numerical results also revealed that all the three effects are in the same direction and add up to about 600 pcm reactivity increase, bringing a good consistency between the multi-group calculation and continuous energy Monte Carlo calculation.© 2013 Elsevier Ltd. All rights reserved. Source

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