scholarly journals Simulation of the burnup in cell calculation using the wimsd-5b code considering different nuclear data libraries

2021 ◽  
Vol 9 (2A) ◽  
Author(s):  
Desirée Yael de Sena Tavares ◽  
Adilson Costa Da Silva ◽  
Zelmo Rodrigues De Lima
Keyword(s):  
Transport Equation ◽  
Transport Theory ◽  
Neutron Transport ◽  
Nuclear Data ◽  
Collision Probability ◽  
Probability Method ◽  
Neutron Transport Equation ◽  
Fuel Burning ◽  
Data Libraries ◽  
Nuclear Data Libraries

This work proposes to implement the cell calculation considering the fuel burning using the WIMSD-5B code. The cell calculation procedure allows determining the nuclear parameters present in the multi-group neutron diffusion equation and for this purpose the neutron transport theory is used in a problem with dimensional reduction, but in contrast is considered a large number of groups associated with the neutron spectrum. There are a variety of reactor physics codes that determine the nuclear parameters by solving the neutron transport equation applied to an equivalent cell representing a fuel element. The WIMSD-5B code is a deterministic code that solves the transport equation using collision probability method. The simulation of fuel burning in the cell calculation took into account different nuclear data libraries. The WIMSD-5B code supports several nuclear data libraries and in the present work the following libraries were used: IAEA, ENDFB-VII.1, JENDL3.2, JEFF3.1 and JEF2.2, all formatted for 69 energy groups.

scholarly journals Solution of neutron-transport multigroup equations system in subcritical systems

2018 ◽  
Vol 4 (1) ◽  
pp. 79-85 ◽  
Author(s):  
Igor V. Shamanin ◽  
Sergey V. Bedenko ◽  
Vladimir N. Nesterov ◽  
Igor O. Lutsik ◽  
Anatoly A. Prets
Keyword(s):  
Neutron Flux ◽  
Transport Equation ◽  
Flux Density ◽  
Diffusion Approximation ◽  
Neutron Transport ◽  
Nuclear Data ◽  
Neutron Transport Equation ◽  
Neutron Flux Density ◽  
Data Libraries ◽  
Nuclear Data Libraries

An iteration method has been implemented to solve a neutron transport equation in a multigroup diffusion approximation. A thermoelectric generator containing plutonium dioxide, used as a source of thermal and electric power in spacecraft, was studied. Neutron yield and multigroup diffusion approximation data was used to obtain a continuous and group distribution of neutron flux density spectra in a subcritical multiplying system. Numerical multigroup approaches were employed using BNAB-78, a system of group constants, and other available evaluated nuclear data libraries (ROSFOND, BROND, BNAB, EXFOR and ENDSF). The functions of neutron distribution in the zero iteration for the system of multigroup equations were obtained by approximating an extensive list of calculated and experimental data offered by the EXFOR and ENDSF nuclear data libraries. The required neutronic functionals were obtained by solving a neutron transport equation in a 28-group diffusion approximation. The calculated data was verified. The approach used is more efficient in terms of computational efforts (the values of the neutron flux density fractions converge in the third iteration). The implemented technique can be used in nuclear and radiation safety problems.

scholarly journals Solution methods of neutron transport equation in nuclear reactors

2013 ◽  
Vol 14 (2) ◽  
pp. 59
Author(s):  
Mohamad Ali Shafii
Keyword(s):  
Monte Carlo ◽  
Transport Equation ◽  
Nuclear Reactor ◽  
Nuclear Reactors ◽  
Neutron Transport ◽  
Collision Probability ◽  
Computer Memory ◽  
Neutron Transport Equation ◽  
Solution Methods ◽  
Transport Approximation

A few numerical methods that usually used to solve neutron transport equation in nuclear reactor are SN dan PN method, Monte Carlo, Collision Probability and Methods of Characteristics . First two methods have been developed using diffusion approach, and last three methods suitable are applicated for transport approximation. Those of three methods have important role in the desain of nuclear reactors. In addition to follow the development of advanced reactor designs, the three methods were chosen because they do not use diffusion approach these are more accurate methods, as well as less need considerable computer memory. Of all the existing methods, the CP method has several advantages among the others. Keywords : Neutron transport, SN, PN, CP, MOC, MC

Determinant Methods for Solving Neutron Transport Equation in Unstructured Geometry

2010 ◽  
Author(s):  
Hongchun Wu ◽  
Guoming Liu ◽  
Liangzhi Cao ◽  
Qichang Chen
Keyword(s):  
Finite Element Method ◽  
Finite Element ◽  
Fuel Assembly ◽  
Transport Equation ◽  
Neutron Transport ◽  
Transmission Probability ◽  
Probability Method ◽  
Neutron Transport Equation ◽  
Nodal Method ◽  
Element Method

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.

scholarly journals The Pij matrix and flux calculation of one-dimensional neutron transport in the slab geometry of nuclear fuel cell using collision probability method

2018 ◽  
Vol 197 ◽  
pp. 02006 ◽  
Author(s):  
Mohammad Ali Shafii ◽  
Jakaria Usman ◽  
Seni H. J. Tongkukut ◽  
Ade Gafar Abdullah
Keyword(s):  
Fuel Cell ◽  
Boundary Conditions ◽  
Neutron Flux ◽  
Nuclear Fuel ◽  
Neutron Transport ◽  
Collision Probability ◽  
Probability Method ◽  
Neutron Transport Equation ◽  
Slab Geometry ◽  
One Dimensional

Calculation of Pij matrix of one-dimensional neutron transport in the slab geometry of the nuclear fuel cell using Collision Probability (CP) method has been done. Pij matrix is one of important parameters within the distribution of neutron flux in the nuclear fuel cell. The CP method is the most efficient methods to solve the neutron transport equation in the reactor core. The study is focused on neutron interaction with nuclear fuel cell of U-235 and U-238 for homogeneous condition. The parameters to calculate the Pij matrix are the cross section of nuclear fuel, width of the region and number of regions. A lattice of slabs have been constructed using void boundary conditions for model of finite system to calculate the collision probabilities. If the Pij matrix has been calculated then neutron flux can be determined. The results show that total value of Pij matrix using CP method for U-235 and U-238 is less than one, respectively. This is in accordance with the definition of void boundary conditions for finite slab geometry. Along with Pij matrix, neutron flux is also appropriate with the reference.

Transmission Probability Method Based on Triangular-Z Mesh for Solving Neutron Transport Equation in Three-Dimensional Geometry

2008 ◽  
Author(s):  
Guoming Liu ◽  
Hongchun Wu
Keyword(s):  
Transport Equation ◽  
Present Method ◽  
Flux Distribution ◽  
Monte Carlo Calculation ◽  
Neutron Transport ◽  
Three Dimensional ◽  
Transmission Probability ◽  
Benchmark Problems ◽  
Probability Method ◽  
Neutron Transport Equation

This paper presents a transmission probability method (TPM) to solve the neutron transport equation in three-dimensional triangular-z geometry. The neutron source within the mesh is assumed to be spatially uniform and isotropic. On the mesh surface, the constant and the simplified P1 approximation are invoked for the anisotropic angular neutron flux distribution. Based on this model, a code TPMTDT is encoded. It was verified by three 3D Takeda benchmark problems, in which the first two problems are in XYZ geometry and the last one is in hexagonal-z geometry. The numerical results of the present method agree well with those of Monte-Carlo calculation method and Spherical Harmonics (PN) method.

High-Dimensional Model Representations for the Neutron Transport Equation

2014 ◽  
Vol 177 (3) ◽  
pp. 350-360 ◽  
Author(s):  
Zhengzheng Hu ◽  
Ralph C. Smith ◽  
Jeffrey Willert ◽  
C. T. Kelley
Keyword(s):  
Transport Equation ◽  
Neutron Transport ◽  
High Dimensional ◽  
Dimensional Model ◽  
Neutron Transport Equation
Sign in / Sign up

Export Citation Format

Share Document