Anisur Rahman
- 16
- 1
- TL;DR
- I am simulating a medical LINAC in MCNP5 and trying to calculate the absorbed dose at the isocenter, but my tally is not producing any meaningful results. I am looking for help identifying the problem with my source, geometry, or tally setup.
I am working on a Monte Carlo simulation of a medical linear accelerator (LINAC) using MCNP5. My objective is to calculate the radiation dose at the isocenter of the LINAC treatment setup.
I have constructed the LINAC geometry, including the relevant components of the treatment head, and defined an electron beam as the source. I have also defined a tally at the isocenter to estimate the dose delivered at that location.
However, after running the MCNP input file, I am not getting any useful tally data for the isocenter. The calculation itself appears to run, but the expected dose/tally result is zero (0).
I am particularly unsure whether the problem is related to:
I am using MCNP5. I have attached the complete input file, including the geometry, source definition, tally cards, and material definitions, if needed.
Could anyone help me identify what might be preventing the isocenter dose tally from producing a result? I would particularly appreciate advice on whether my tally setup is appropriate for calculating dose at the isocenter in a LINAC simulation.
I have constructed the LINAC geometry, including the relevant components of the treatment head, and defined an electron beam as the source. I have also defined a tally at the isocenter to estimate the dose delivered at that location.
However, after running the MCNP input file, I am not getting any useful tally data for the isocenter. The calculation itself appears to run, but the expected dose/tally result is zero (0).
I am particularly unsure whether the problem is related to:
- The definition or placement of the tally at the isocenter
- The source definition and electron beam parameters
- Particle transport through the LINAC head
- The choice of tally for calculating dose
- The energy binning or DE/DF cards
- The normalization of the tally
- The geometry/material definitions
- Particle importances or particle cutoffs
I am using MCNP5. I have attached the complete input file, including the geometry, source definition, tally cards, and material definitions, if needed.
Could anyone help me identify what might be preventing the isocenter dose tally from producing a result? I would particularly appreciate advice on whether my tally setup is appropriate for calculating dose at the isocenter in a LINAC simulation.