Abstract
An in-house computer code MANTRA is developed to solve the neutron transport equation at lattice or assembly level (in 2D) applying Method of Characteristics (MOC) for reactor design calculation. The code has the capability to divide the lattice into triangular meshes. By default, the code assumes constant neutron source inside the meshes. If required, it is possible to use linear source assumption to reduce the number of meshes for speeding up the solution. Using MOC, the solution of transport equation is obtained along a number of parallel lines or rays traced through the meshes in different directions and connect them through reflective or periodic boundary condition. In the code, there is option to choose either Power or Krylov subspace iteration technique to converge the solution with desired accuracy. The code is coupled with multigroup cross section library WIMSD. Performance of MANTRA is tested against a number of benchmark problems and overall good agreement is observed with the results obtained from internationally recognized codes like WIMSD, DRAGON, MCNP etc.
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