. It shows error in the run section .
42- sdef erg=100 par=1 sur=80 ext=0 0 1.96 rad=0.47
warning. ext is constant. in most problems it is a variable.
fatal error. illegal entry: 0f2:n 111.1
e2 0 224i 101
nps 30000No dear , this is exactly the output what I got . I recheck after uploading. This is also a matter of sadness for me for few days !!!Alex A said:
actually there was issues with my laptop , code runs here but did not produce any data. I run it to another laptop, it give me data, but that was not enough for getting spectra,Alex A said:
I modified the output as of your codePSRB191921 said:
To set up a basic neutron source in MCNP, you need to define the source particles, their energy distribution, and spatial distribution. This is typically done using the SDEF (Source Definition) card. For example, to define a monoenergetic point source, you might use:
SDEF POS=0 0 0 ERG=2.45This sets the source at the origin (0,0,0) with an energy of 2.45 MeV. You can customize the source further by specifying different distributions and parameters as needed for your experiment.
MCNP offers several tally options for neutron spectroscopy, including F4 (flux tally), F5 (point detector tally), and F8 (energy deposition tally). For neutron spectroscopy, the F4 tally is commonly used to measure neutron flux in a cell, and the F8 tally can be used to measure energy deposition, which can be correlated with neutron spectra. To use an F4 tally, you might include:
F4:N 1where 'N' indicates neutrons and '1' is the cell number. You can refine this tally with energy bins using the E4 card to get spectral information.
Modeling a detector in MCNP involves defining the geometry of the detector, the materials it is made of, and the appropriate tallies to record the interactions. For instance, if you are using a helium-3 detector, you would define the detector region with helium-3 gas and use an F8 tally to measure the energy deposition. You might set up the geometry and tally like this:
CELL 1 0 -1 IMP:N=1CELL 2 1 -2 IMP:N=1MATERIAL 1 2004.70c 1.0F8:N 2This sets up a simple helium-3 detector and uses an F8 tally to record the neutron interactions within the detector cell.
Common sources of error in neutron spectroscopy simulations include incorrect source definitions, improper tally settings, and inadequate statistical sampling. To minimize these errors, ensure your source definition accurately represents the physical source, verify that your tallies are correctly set up to capture the desired data, and run sufficient particles to achieve statistically significant results. Additionally, using variance reduction techniques such as importance sampling and weight windows can help improve the efficiency and accuracy of your simulations
