Recent content by PSRB191921
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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...- PSRB191921
- Post #4
- Forum: Biology and Medical
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Dose rate measurement through MCNP
mcnp gives the result for one particle. Simply multiply by the number of particles emitted per second- PSRB191921
- Post #4
- Forum: Nuclear Engineering
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Dose rate measurement through MCNP
Hi, 1 Gy = 1 J/kg 1 eV = 1.6022E-19 J hope it's help- PSRB191921
- Post #2
- Forum: Nuclear Engineering
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News 2024 Summer Olympic Games Discussion
let's remember this- PSRB191921
- Post #13
- Forum: General Discussion
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News 2024 Summer Olympic Games Discussion
A man? No a football legend! Sorry a soccer legend!- PSRB191921
- Post #11
- Forum: General Discussion
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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...- PSRB191921
- Post #2
- Forum: Engineering and Comp Sci Homework Help
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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.- PSRB191921
- Post #3
- Forum: Nuclear Engineering
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What Could Be Causing Unexpected Results in the LET Analysis with MCNP6.2?
oups you are right !- PSRB191921
- Post #5
- Forum: Nuclear Engineering
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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- PSRB191921
- Post #3
- Forum: Nuclear Engineering
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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.- PSRB191921
- Post #17
- Forum: Nuclear Engineering
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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...- PSRB191921
- Post #14
- Forum: Nuclear Engineering
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Calculating Microdosimetry in MCNP
this is my Excel file- PSRB191921
- Post #15
- Forum: Nuclear Engineering
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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)- PSRB191921
- Post #13
- Forum: Nuclear Engineering
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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...- PSRB191921
- Post #12
- Forum: Nuclear Engineering
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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 ?- PSRB191921
- Post #11
- Forum: Nuclear Engineering