Determine the intensity of fission gamma

[Xue]

Hello,

I want to calculate the photon dose rate outside a reactor when it is at full power, using MAVRIC.

I plan to use the build-in gamma energy spectrum to sample the gamma ray energy.

distribution 2 special="fissionPhotons"  parameters   92235 end end distribution

However, I am not sure how to calculate the total photon intensity (gamma/second) to define the photon source.

I can estimate the fission reactions/second by dividing the thermal power by 190MeV/fission. Unlike the v value for fission neutrons, I do not know on average how many photons are emitted per fission reaction. 

From textbook, each fission releases 7 MeV prompt gamma and 6.3 MeV decay gamma. I can approximate this value as 13.3 gamma rays of 1 MeV each. 

i wonder if there is a better approach for more accurate results.

Thanks for your valuable advise. 

Xue

[Cihangir]

Dear Xue,

You can use the special distributions "fissionPhotons" to generate the energy distribution of fission photons, but beware that option is only supported with multigroup libraries. You can run a multi-group library to get the distribution and use it in a continuous energy simulation later. If you are looking for a more detailed distribution, then you can get point-wise fission energy distributions and multiplicities from the ENDF libraries for specific nuclides; try NNDC (https://www.nndc.bnl.gov). On average, 4 to 7 fission gammas are emitted per each fission event with energies up to 8 MeV.

As a better alternative, I recommend generating the fission source for your model using CSAS by enabling "cds" option in the parameters and importing the fission source into the MAVRIC by converting generated mesh tally (#DMAP file) to a mesh source (MSM file). There is a MAVRIC utility to do that; look up "mt2msm" in the manual. You can use the number of fissions you calculate based on the thermal power, as you mentioned in your post. MAVRIC will use both the nubar to be used as the multiplicity for neutrons and also generate fission gammas even if the fissions are disabled in the model (fissionMult=0) as long as secondary gammas are enabled (secondaryMult>0). This method not only accounts for fission neutrons and gammas but also includes secondary gammas due to neutron interactions that you would ignore if you were only modeling the fission gammas.

Cheers,

Cihangir

[Xue]

Hello Cihangir,

Thanks for your reply. I have a few more questions.

I think the 3dmap file only contains the fission neutron source information, not the fission gammas. Do you mean that by using a fission neutron source, the fission gamma can also be produced? 

I tend to use the built-in gamma spectrum. I previously used the MSM file as the fission neutron source for neutron dose calculation, but it always crashes, saying the geometry is wrong (undefined or doubly defined). I think my geometry is correct, and I later changed it to the built-in fission neutron spectrum without using the MSM file, and it works well. 

Based on your suggestion, I compiled the following partial input to calculate the dose rate from fission gammas, excluding the secondary gamma generated from fission neutrons. 

In the strength,  this input uses the (fission neutrons/sec) in the core, that is (core thermal power)/(190MeV/fission) * nubar. The source type is set to neutron with fission neutron energy and space distribution, while the response is set to photons.

As expected, the photon response is zero, because no photon is ever generated. I think this approach can only calculate the secondary gamma from fission neutrons, but not the dose rate from the fission gamma, if I turn on the secondarymulti=1. 

Would you help me to revise it? Thanks. 

' Response on page 2353, flux-2-dose on 2355. Photon:9505, neutron:9031

' 50mSv/year = 5rem/year = 5/2080 = 2.4e-3 rem/h

' response 1

' title="ANSI standard (1991) flux-to-dose-rate factors for neutrons, rem/h"

' doseData=9031 noExtrapolation end response

response 2

title="ANSI standard (1991) flux-to-dose-rate factors for photons, rem/h"

doseData=9505 noExtrapolation end response

distribution 1 special="fissionNeutrons" parameters 1 92235 end end distribution

distribution 2 special="fissionPhotons" parameters 92235 end end distribution

distribution 101 special="pwrNeutronAxialProfile" end distribution

distribution 102 special="pwrGammaAxialProfile" end distribution

end definitions

' ============================

' Source block, on page 2367

' ============================

read sources

'Total neutron intensity: thermalPower/(193.7MeV/fission) * nubar

'Total neutron intensity: 83.3 MW / (193.7MeV/fission) * 2.624979 = 7.045798e18 neutrons/second

src 1

title='Fission source'

strength=7.045798e18 neutron eDistributionID=1

cuboid 42.83456 -42.83456 42.83456 -42.83456 182.88 -182.88

zDistributionID=101 zScaleDist

mixture=1 end src

end sources

' ============================

' Tallies block, on page 2371

' ============================

' Detector definition

read tallies

pointDetector 11 title="Photons: point detector, side"

locationID=11 responseID=2 photon end pointDetector

' pointDetector 12 title="NEUTRONS: point detector, side"

' locationID=12 responseID=1 neutrons end pointDetector

' pointDetector 13 title="NEUTRONS: point detector, side"

' locationID=13 responseID=1 neutrons end pointDetector

' pointDetector 14 title="NEUTRONS: point detector, side"

' locationID=14 responseID=1 neutrons end pointDetector

' pointDetector 21 title="NEUTRONS: point detector, roof"

' locationID=21 responseID=1 neutrons end pointDetector

' pointDetector 22 title="NEUTRONS: point detector, roof"

' locationID=22 responseID=1 neutrons end pointDetector

end tallies

read parameters

randomSeed=00003ecd7b4e3e8e

library="v7-200n47g"

perBatch=1000000 batches=100

fissionMult=0

secondaryMult=0

' noPhotons: turn off the secondary photon generation from neutron interactions

' noPhotons

end parameters

[Cihangir]

Xue,

The 3DMAP from CSAS calculations only include spatial and energy distributions of fission neutrons and have nothing about gammas, as you pointed out. MAVRIC generates fission gammas when transporting neutrons from the imported MSM files if you have enabled gammas in the parameters block (secondaryMult>0 and photons). If you want to use the MG libraries with built-in  "fissionNeutrons" and "fissionPhotons" (both are for only energy distributions) then you should define both neutron and gamma sources in your source block. You should define source strengths based on your reactor's thermal power and particle multiplicity, so use nubar for the fission neutrons and 4 to 7 (a single value can be obtained by folding the energy spectrum with the multiplicity) for the fission gammas.

Thanks,

Cihangir

[Xue]

Hello Cihangir,

Thanks a lot for your advice. 

I see the difference between the MSM and fissionPhotons. 

I will do some tests and update in a few weeks. 

Sincerely,