Recent content by PSRB191921

  1. PSRB191921

    Neutron contamination threshold in tissue using LINAC

    An old IEC standard (International Electrotechnical Commission) has proposed a maximum neutron dose limit in the patient plane of 0.5 mGy of neutrons per Gy of x-ray. I don't know if this is still relevant. In addition in term of energy, you have a large number of photonuclear reaction...
  2. PSRB191921

    Dose rate measurement through MCNP

    mcnp gives the result for one particle. Simply multiply by the number of particles emitted per second
  3. PSRB191921

    Dose rate measurement through MCNP

    Hi, 1 Gy = 1 J/kg 1 eV = 1.6022E-19 J hope it's help
  4. PSRB191921

    News 2024 Summer Olympic Games Discussion

    let's remember this
  5. PSRB191921

    News 2024 Summer Olympic Games Discussion

    A man? No a football legend! Sorry a soccer legend!
  6. PSRB191921

    MCNP Help: 10 Particles Lost

    Hi, you are trying to calculate an F5 (and F15) at (0,0,0) where your neutrons are emitted. Try to remove this calculation point, like this : F5:N 10 0 0 1 20 0 0 1 30 0 0 1 40 0 0 1 50 0 0 1 60 0 0 1 70 0 0 1 80 0 0 1 90 0 0 1 100 0 0 1...
  7. PSRB191921

    I need help in solving the problem of the code written with MCNPX 2.6

    Hi, To calculate a spectrum in Ge, the tally F8 is the right tally. In the MCNP pack I don't think there is a proton library for B-11, so a model is applied (you can look in your xsdir or your output file). I don't know if this is the problem but you should try installing a p-B11 library to try.
  8. PSRB191921

    What Could Be Causing Unexpected Results in the LET Analysis with MCNP6.2?

    Hi, The your result is logical. 1E-5 is in MeV/cm i.e. 100 keV/µm but for electrons the LET is much lower than this value: all results are in the bin 0 - 1E-5 therefore also between 0 and 1e-3. you should sample between 0 and less than 10 keV/µm
  9. PSRB191921

    MCNP code for Neutron Spectroscopy

    Am-Be is not a fission source, so it is not a watt spectrum. You must input the spectrum by bin.
  10. PSRB191921

    MCNP code for Neutron Spectroscopy

    with F5:n x y z .1 you calculate the fluence at the coordinate (x,y,z). With DE/DF the fluence is transformed into equivalent dose. In my input file I put : F5:n 0 0 10 .1 for a distance of 10 cm from the source F15:n 0 0 50 .1 for a distance of 50 cm from the source F25:n 0 0 100 .1 for a...
  11. PSRB191921

    Calculating Microdosimetry in MCNP

    this is my Excel file
  12. PSRB191921

    Calculating Microdosimetry in MCNP

    with your file and my processing it gives: not so bad! Some convergence problem (with more nps it will be ok)
  13. PSRB191921

    MCNP code for Neutron Spectroscopy

    your dose calculation is strange, because: - you do the calculation through a sphere while the notion of dose is punctual. In principle with mcnp we calculate the fluence at a point (for example with a type 5 tally) and we apply a DE/DF to it. - Cf-252 is a spectrum not monoenergetic at 2.26 MeV...
  14. PSRB191921

    Calculating Microdosimetry in MCNP

    I don't know your processing code, can you give your output file I will try with my processing code (Excel :-)) I think your curve is for alpha, can your try with protons and alpha+ protons ?
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