Mavric: Calculating absorbed energy (dose) in any material (non tissue)

[Peter Wolniewicz]

I would like to calculate the absorbed energy in for example concrete and plastic from photon/neutron radiation. The built in Flux-to-Dose conversion factor MT numbers, like table 8.2.5 in the manual for Scale 6.2.3 have a few non-Sv/h options, like MT 9503 ( Claiborne-Trubey conversion factors ), but I think these are for human tissue as well.

Is it possible to calculate the absorbed energy in Mavric in any material from photons and neutrons? Perhaps I should use some other ENDF MT response number?

Best regards,

Peter

[Cihangir]

Hi Peter,

Unfortunately, SCALE doesn't provide a built-in MT number for energy absorption/deposition in MAVRIC. You can ask for any available reaction MT number in a response (for both a nuclide or a material) to be used but it will be up to the users to select what to use. Additionally, I recommend doing a literature search for your specific materials and possibly look at IRDF reactions and verified/validated KERMA data sets.

Thanks,

Cihangir

[Peter Wolniewicz]

Cihangir, thank you for the advice! I have little knowledge of what combo of MT number to use here but I have asked the same question on the MCNP forum.

[Peter Wolniewicz]

Just to make sure that I understand:

There is no way to calculate the energy deposition in a material in any way with the SCALE package in order to produce [MeV/g]?

I was thinking of something like the F6 or *F6 tally in MCNP6 ("the Track length estimate of energy deposition").

Best regards,

Peter 

[Cihangir]

SCALE doesn't have a tally for energy deposition, which is F6 in MCNP. Instead, available flux-to-dose-conversion factors or user responses can be used as a multiplier to tallies (track-length estimators), an indirect way of two-step energy deposition tally.

[Peter Wolniewicz]

Could you give an example of such a tally or is it material specific?

What I have is a .f71 file with all sorts of nuclides emitting alpha, beta, photons.
I import this to mavric.

Tally certain area.

Thank you, and I apologize for all the questions. I understand if the answer is it's possible but hard to do :)

Best regards Peter

[Cihangir]

ORIGIN f71 file only provides the emission energy distribution for your sources in MAVRIC. 

Assume that you are tallying a region (in unit 1 and region 1) made of pure iron-56 and trying to calculate radiative capture heating.

Your response and tally definitions then would look like:

    response 1

        title="Neutron radiative capture heating  for Fe-56 (eV-barns)"

        neutron

        bounds

             2.0e+07 1.0e+07 1.0e+06 1.0e+05 1.0e+04 1.0e+03 1.0e+02 1.0e+01 

             1.0e+00 1.0e-01 1.0e-02 1.0e-03 1.0e-04 1.0e-05 end

        values   

             8.1e-01 2.6e-01 5.0e-01 1.4e-02 5.5e-01 2.4e+00 7.7e+00 2.5e+01 

             7.8e+01 2.5e+02 7.8e+02 2.5e+03 7.8e+03 2.5e+04 end

    end response

    regionTally 1

        title="Iron heating (eV/s)"

        neutron

        unit=1 region=1

        responseID=1 

              '    convert the unit of response to eV-cm**2

                   multiplier=1e-24

              '    factor for converting tally to total heating in the region, assuming the volume of 1 cm3 and density of 8 g/cc

                   multiplier=1*8/56

    end regionTally