Transmission Probability Method of Integral Neutron Transport Calculation for Two-Dimensional Rectangular Cells

1975 ◽  
Vol 56 (4) ◽  
pp. 411-422 ◽  
Author(s):  
Hans Häggblom ◽  
Åke Ahlin ◽  
Takashi Nakamura
Author(s):  
Liang Liang ◽  
Hongchun Wu ◽  
Liangzhi Cao ◽  
Youqi Zheng

The method of characteristics (MOC) has been widely used in lattice code for its high precision and easy complement. However, the long characteristics method needs large quantity of PC memory when dealing with large scale problems. The modularity MOC method could significantly reduce the PC memory when calculating the problem which contains lots of repeatedly geometries, like the fuel assembly in the reactor. In this method, only typical geometric cells are selected to trace the rays, and then the geometry information of these cells is stored. So, the modularity MOC method is feasible to perform well in the calculation with large scale. When tracing the rays, the technique of mesh ray generating and the corresponding azimuthal quadrature set are both applied. The techniques make sure that each ray has the reflected ray in the boundary so it is convenient to describe the boundary condition. The optimal polar angle and the Guass quadrature set are selected as the polar quadrature set. Furthermore, the coarse mesh finite difference (CMFD) is employed to accelerate the calculation. A pin cell is chosen as the coarse mesh. The CMFD solution provides the MOC with much faster converged fission and scattering source distributions. The LOTUS code is developed and the numerical results show that the code is precise for engineering application and the CMFD acceleration is effective.


Author(s):  
Hongchun Wu ◽  
Guoming Liu ◽  
Liangzhi Cao ◽  
Qichang Chen

The spherical harmonics (Pn) finite element method, the Sn finite element method, the triangle transmission probability method and the discrete triangle nodal method were all introduced to solve the neutron transport equation for unstructured fuel assembly respectively. The computing codes of each method were encoded and numerical results were discussed and compared. It was demonstrated that these four methods can solve neutron transport equations with unstructured-meshes very effectively and correctly, they can be used to solve unstructured fuel assembly problem.


2021 ◽  
Vol 247 ◽  
pp. 04011
Author(s):  
Yasushi Nauchi ◽  
Alexis Jinaphanh ◽  
Andrea Zoia

Time-dependent neutron transport in non-critical state can be expressed by the natural mode equation. In order to estimate the dominant eigenvalue and eigenfunction of the natural mode, CEA had extended the α-k method and developed the generalized iterated fission probability method (G-IFP) in the TRIPOLI-4® code. CRIEPI has chosen to compute those quantities by a time-dependent neutron transport calculation, and has thus developed a time-dependent neutron transport technique based on k-power iteration (TDPI) in MCNP-5. In this work, we compare the two approaches by computing the dominant eigenvalue and the direct and adjoint eigenfunctions for the CROCUS benchmark. The model has previously been qualified for keffs and kinetic parameters by TRIPOLI-4 and MCNP-5. The eigenvalues of the natural mode equations by α-k and TDPI are in good agreement with each other, and closely follow those predicted by the inhour equation. Neutron spectra and spatial distributions (flux and fission neutron emission) obtained by the two methods are also in good agreement. Similar results are also obtained for the adjoint fundamental eigenfunctions. These findings substantiate the coherence of both calculation strategies for natural mode.


2020 ◽  
Vol 8 ◽  
Author(s):  
Peitao Song ◽  
Qian Zhang ◽  
Liang Liang ◽  
Zhijian Zhang ◽  
Qiang Zhao

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