Time filter

Source Type

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.


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.


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.


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.


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.


Han Y.,Shanghai JiaoTong University | Jiang X.-F.,Shanghai NuStar Nuclear Power Technology Co. | Wang D.-Z.,Shanghai JiaoTong University
Nuclear Science and Techniques | Year: 2016

An improvement for application of Dancoff factor is developed. It combines Stamm'ler's two-term method for resonance integral calculation with neutron current method for Dancoff factor calculation. Stamm'ler's formulation, which is originally derived for the infinite lattice geometry, can be easily revised to contain the Dancoff factor explicitly, while the neutron current method can easily calculate the Dancoff factor for general irregular assembly geometry. For the resonance interference effects, the resonance interference factor table is built in pairs of nuclides, only for the interference between 238U and other resonance nuclides, spanning over a range of background cross-section and number density ratio of the pairing nuclides. A series of verification calculations have been carried out for problems of infinite lattice and single assembly geometry, with two or multiple resonance absorbers. For these verification calculations, our improvement on Dancoff factor application and resonance interference give good results. © Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Chinese Nuclear Society, Science Press China and Springer Science+Business Media Singapore 2016.


Zhang S.,Shanghai NuStar Nuclear Power Technology Co. | Chen G.,Shanghai NuStar Nuclear Power Technology Co.
International Conference on Mathematics and Computational Methods Applied to Nuclear Science and Engineering, M and C 2013 | Year: 2013

This paper presents a description of methodologies and preliminary verification results of a new lattice physics code ROBIN, being developed for PWR application at Shanghai NuStar Nuclear Power Technology Co., Ltd. The methods used in ROBIN to fulfill various tasks of lattice physics analysis are an integration of historical methods and new methods that came into being very recently. Not only these methods like equivalence theory for resonance treatment and method of characteristics for neutron transport calculation are adopted, as they are applied in many of today's production-level LWR lattice codes, but also very useful new methods like the enhanced neutron current method for Dancoff correction in large and complicated geometry and the log linear rate constant power depletion method for Gd-bearing fuel are implemented in the code. A small sample of verification results are provided to illustrate the type of accuracy achievable using ROBIN. It is demonstrated that ROBIN is capable of satisfying most of the needs for PWR lattice analysis and has the potential to become a production quality code in the future.


Chen G.,Shanghai Nustar Nuclear Power Technology Co. | Huang Y.,Shanghai Nustar Nuclear Power Technology Co. | Jiang X.,Shanghai Nustar Nuclear Power Technology Co. | Wang T.,Shanghai Nustar Nuclear Power Technology Co. | Zhang S.,Shanghai Nustar Nuclear Power Technology Co.
Hedongli Gongcheng/Nuclear Power Engineering | Year: 2014

Verification and validation of lattice code ROBIN-1.7 is performed by using international public benchmark problems, which include critical experiments data, depletion benchmark problems of OECD NEA and other neutron transport-depletion calculations with MCNP code. The result (reactivity, pin power distribution, isotope concentration) shows that independent modules of ROBIN-1.7 such as resonance treatment module, neutron transport module and depletion module are developed and integrated correctly. It is also demonstrated that ROBIN-1.7 has reached the industry level for PWR application. ©, 2014, Yuan Zi Neng Chuban She. All right reserved.


Wang T.,Shanghai NuStar Nuclear Power Technology Co. | Jiang X.,Shanghai NuStar Nuclear Power Technology Co. | Lyu D.,Shanghai NuStar Nuclear Power Technology Co. | Chen G.,Shanghai NuStar Nuclear Power Technology Co. | And 2 more authors.
Hedongli Gongcheng/Nuclear Power Engineering | Year: 2014

Code validations are performed by comparing the calculated results against the measured data from three types of operating reactors. A total of 59 reactor cycles are evaluated and the results compared not only include the parameters measured at the startup physics test stage for zero-and low-power conditions but also that measured at the normal operation stage for full-power conditions. Validation results demonstrate that the neutronic models incorporated in ORIENT system are of high quality and the system is highly acceptable for performing routine neutronic analyses at PWR nuclear power plants. ©, 2014, Yuan Zi Neng Chuban She. All right reserved.

Loading Shanghai NuStar Nuclear Power Technology Co. collaborators
Loading Shanghai NuStar Nuclear Power Technology Co. collaborators