Recent content by damyoro

Help for ?CNP input Greetings I am using MCNP to estimate the shielding of Cobalt source placed at the center of cube of 2 meter of side. I wrote an input file using macroby. I used successive cubes and filled them with either air or lead (or other shielding material: tungsten, steal, etc) to...

Greetings to all I need some some help to proceed the calculations of gamma shielding using MCNP. The details of the problem are as follows: - A square irradiation room/cell of maximum of 2 m of side - A cobalt source with an activity of 5,000 Ci placed on a table inside the room, - A glass...

Thank you for your help and your availability. I really appreciated. I will focus on reading the manual.

If I got it right, here is the code MNCP Input for Concrete Attenuation calculation c c Cell cards: 1 1 -2.3 -1 5 1 -2.3 -5 10 1 -2.3 -10 15 1 -2.3 -15 20 2 -0.001205 -20 25 2 -0.001205 -25 30 2 -0.001205 -30 35 2 -0.001205 -35 40 2 -0.001205 -40 45 2 -0.001205 -45 c Tally... 60 2 -0.001205 -60...

I use both MCNP5 visual Editor and command line. When I run the file the nps count doesn't even start. 1. This is the message generated when I run with Visual inp = cc4.txt outp = CC4O runtpe = run mctal = runtal starting mcnp execution comment. 22 surfaces were deleted for being...

Input file MCNP5 Greetings to all After some corrections, the MCNP input files for shielding calculations for iron and lead run without problem, but the input file for the concrete shielding calculations did not work, although the files are similar , I have only replaced in the material...

Thank you very much for your helpful reply. I will certainly follow your advice and come back to you soon.

Greetings to all I had some problems using MCNP5 for gamma shielding calculation. The original input file is from a document by George E. Chabot, Jr., PhD, CHP Shielding of Gamma Radiation. The description of the problems are following: 1) Three different shielding materials, namely...

You are right and thank you for your reply This is my try general solution is C=A*e^(-x/L)+Be^(x/L) And I use as boundary conditions φ(a/2)==0 and as source condition J(a/2-x')=J(x') +1 What I found is the following Ф+(x,x^' )=-(Le^((x-a/2)/L))/2D[cosh x^'/L-cosh((x^'-a/2)/L) ]...

I wanted to add the reference of the book : Introduction to Nuclear reactor theory by John R. Lamarsh CHp5 Problem a)By using the diffusion equation for a planar source located at x', show that the diffusion kernel for an infinite slab of thickness a is given by G(x,x^' )=L/(Dsinh(a/L))...

Deaar all good morning I am very interested to the flux in a slab of extrapolated thickness a, containing distributed sources of neutron. A I have an example in which the source is given as s(x)=S(x+a/2) where S is a constant and x distance from the center of the slab. You mentioned in one...