Computer Assignment No. 1
Due March 31, 2017
NEEP 408
MWK
1. A monodirectional beam of neutrons is incident on a slab and is composed of prompt fission
neutrons. The incident radiation is produced by a fission plate having a fission rate of 2.0 x
1012 fiss/cm2 -s. Determine the thickness of shield material required to reduce the biological
contact dose rate to the allowable average weekly dose rate limit for a radiation worker [2.5
mrem/hr] using the DANTSYS code for shields composed of H2 O, SiC (ρ= 3.217 g/cm3 ),
molar mass = 40.10 grams/mole), tungsten (W), two concretes: ordinary portland (ρ= 2.3
g/cm3 ) and iron portland (ρ=5.9 g/cm3 ), and one material of your interest. The concrete
elemental compositions can be found in Vol. 2, Chapter 9 of the reference text entitled “Engineering Compendium on Radiation Shielding” which has been placed in the course book
reserve at Steenbock library. Note, also that there are several isotopes of tungsten listed as
options in the slab neutron file. You must mix / make natural tungsten by using the natural
abundances from the individual isotopes.
Hand in: a) A small table summarizing the comparison of the thicknesses and associated
dose rates for each of the shield materials.
b) a semi-log plot of your results (Dose vs. Distance (all curves on one plot)) and discuss
them (which shield material is the most effective and why).
c) Also generate a spectrum plot (Neutron Flux vs. Energy) for each of the shields. Choose
three points within the shield and label each curve with the corresponding position in the
shield. Hand in the plots, a table summarizing the thickness results and a table with the
number densities used. In addition hand in your dantsys and doseplt input files for the two
concrete cases.
Source calculation:
The incident radiation is produced by a fission plate having a fission rate of 2.0 x 1012
fiss/cm2 -s. You need to consult Chapter 3 of the Lamarsh text ”Introduction to Nuclear
Engineering” or equivalent nuclear engineering text (”Nuclear Reactor Analysis” by Duderstadt, Chapter 2) for the energy dependent 235 U Chi-fission spectrum expression. The
average number of neutrons released per fission for 235 U is ν = 2.4. The source energy
distribution must be entered group-wise (group index - g) and is computed as follows:
Fission rate
Sg =
×ν×
0.0178571429
Z Eg−high
χ(E) dE
Eg−low
where χ(E) is the energy dependent expression for the Chi-fission spectrum and Eg−high and
Eg−low are the upper and lower energies of the group boundaries (consult the group structure handout). The group entries are entered as 3r 0.0 X.XXX+XX where X.XXX+XX or
X.XXXeXX is the Sg value computed for the g’th group from the expression above (note X
denotes a numerical integer).
Note: for this problem you must either delete the norm= X.XXeXX entry from your input
file or place a back-slash at the beginning of the line to make it a comment statement. (Also
remember not to exceed the 80 column limit and not use the tab key).