TRIDENT: a two-dimensional, multigroup, triangular mesh discrete ordinates, explicit neutron transport code

In the last decade Vilhena and coworkers10 reported an analytical solution to the two-dimensional nodal discrete-ordinates approximations of the neutron transport equation in a convex domain. The key feature of these works was the application of the combined collocation method of the angular variable and nodal approach in the spatial variables. By nodal approach we mean the transverse integration of the SN equations. This procedure leads to a set of one-dimensional SN equations for the average angular fluxes in the variables x and y. These equations were solved by the old version of the LTSN method9, which consists in the application of the Laplace transform to the set of nodal SN equations and solution of the resulting linear system by symbolic computation. It is important to recall that this procedure allow us to increase N the order of SN up to 16. To overcome this drawback we step forward performing a spectral painstaking analysis of the nodal SN equations for N up to 16 and we begin the convergence of the SN nodal equations defining an error for the angular flux and estimating the error in terms of the truncation error of the quadrature approximations of the integral term. Furthermore, we compare numerical results of this approach with those of other techniques used to solve the two-dimensional discrete approximations of the neutron transport equation6.

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Gemma: Development of a two-dimensional neutron transport code with the method of characteristics

Annals of Nuclear Energy ◽

10.1016/j.anucene.2018.01.043 ◽

2018 ◽

Vol 115 ◽

pp. 367-376

Author(s):

Guillermo Ibarra Reyes ◽

Samuel Vargas Escamilla ◽

Armando Miguel Gómez Torres ◽

Edmundo del Valle Gallegos

Keyword(s):

Method Of Characteristics ◽

Neutron Transport ◽

Two Dimensional ◽

The Method Of Characteristics ◽

Transport Code

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Evaluation of PWR pressure vessel fast neutron fluence benchmarks from NUREG/CR-6115 with ares transport code

Nuclear Technology and Radiation Protection ◽

10.2298/ntrp1703204z ◽

2017 ◽

Vol 32 (3) ◽

pp. 204-210 ◽

Cited By ~ 2

Author(s):

Liang Zhang ◽

Bin Zhang ◽

Cong Liu ◽

Yixue Chen

Keyword(s):

Fast Neutron ◽

Pressure Vessel ◽

Neutron Fluence ◽

Neutron Transport ◽

Source Distribution ◽

Discrete Ordinates ◽

Source Modeling ◽

Transport Code ◽

Vessel Integrity ◽

Angular Discretization

An accurate evaluation of PWR pressure vessel fast neutron fluence is essential to ensure pressure vessel integrity over the design lifetime. The discrete ordinates method is one of the main methods to treat such problems. In this paper, evaluations have been performed for three PWR benchmarks described in NUREG/CR-6115 using ARES transport code. The calculated results were compared to the reference values and a satisfactory agreement was obtained. In addition, the effects of SN numeric and source distribution modeling for pressure vessel fast neutron fluence calculation are investigated. Based on the fine enough grids adopted, the different spatial and angular discretization introduces derivations less than 3 %, and fix-up for negative scattering source causes no noticeable effects when calculating pressure vessel fast neutron fluence. However, the discrepancy of assembly-wise and pin-wise source modeling for peripheral assemblies reaches ~20 %, which indicates that pin-wise modeling for peripheral assemblies is essential. These results provide guidelines for pressure vessel fast neutron fluence calculation and demonstrate that the ARES transport code is capable of performing neutron transport calculations for evaluating PWR pressure vessel fast neutron fluence.

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