I simulate a reactor in MCNP. However I want to know the spectrum (energy) and flux in a specific surface (or cell) away from the nucleus (source). I thought using F2 tally, but I do not want the normilize value. I want the measured integral value. How do I do this?
MCNP (Monte Carlo N-Particle Transport) is used to develop a versatile and accurate full model of the TRIGA reactor. Continuous-energy, generalized-geometry, time-dependent, Monte Carlo radiation transport code designed to track many particle types over broad ranges of energies. The spectral flux is a useful measure for distinguishing signals whose spectrum changes slowly from signals whose spectrum changes quickly.
https://www.google.com/url?sa=i&url=https%3A%2F%2Fwww.researchgate.net%2Ffigure%2FTypical-neutron-flux-spectra-for-different-reactors-types-Fusion-Fast-breeder-High_fig1_319674489&psig=AOvVaw1BaGyV_UAmMLctYUderkuk&ust=1704602477012000&source=images&cd=vfe&opi=89978449&ved=0CBMQjRxqFwoTCJDl3p35x4MDFQAAAAAdAAAAABAD
Hi Otto,
I think F4 tally will work for you. This gives the neutrons/(cm2-source particle). To get the neutrons/cm2-sec in a KCODE calc you would do this:
F4 x Source neutron/fission (aka nubar) x fission/ energy released (eg 193 MeV) x Power (MJ/sec) x 1/keff.
nubar is calculated by MCNP in a kcode calc.
I hope this helps,
dimitris
Dimitrios Kontogeorgakos Hi!
Thank you for the answer. I am more interested in obtain the spectrum. For example, a graphic like the maxwellian spectrum (attached). It doesn't have to be the same, but similar. In short, I need energy and count data, so I can plot a graph of count (particle density) x energy. I did a grid of detectors, but I dont know what I do now, what commands I can use, etc.
Hi Otto Haubrich,
have you tried to used tally energy cards? Basically, the entries in an "En" card are the upper bounds in MeV of the energy bins for the "n" tally. Combined with the F4 card you will get the flux per energy bin. The neutron spectrum depends of what materials the neutrons pass before hitting the target.
Dimitrios Kontogeorgakos I used the ssw and ssr card. I recorded the neutrons coming from the reactor nucleus with the ssw card, using a kcode as my source. After that I used the ssr card in another surface, but the message in prompt was " bank is full. bank backup file is being created. nps=1".
the code enters an infinite loop.
Do you know something about that?
Hi Otto Haubrich ,
Maybe you can try DBCN J 26R x , where x is the bank size. From the MCNP manual (either 6.3 or 6.2) there is this note "DEFAULTs vary by application: x28=2048 for most fixed-source problems, x28=128 for criticality problems, x28=16384 for high-energy problems)".
Do some testing with a larger bank size and with a smaller NPS.
Regards,
dimitris
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