Brachytherapy source specifications

  • Thread starter Thread starter saidhorion
  • Start date Start date
  • Tags Tags
    Source
AI Thread Summary
The discussion revolves around troubleshooting an error encountered while using MCNPX for calculating energy deposition around a cobalt source. The user reports an issue with the NPS parameter, specifically receiving an error stating "all entries must be integers" when attempting to set NPS to 3E+9. Another participant notes that since version 2.5.0, MCNPX supports 64-bit operation, allowing for up to 2E18 particles. The user confirms they are using version 2.7 on a 64-bit Windows 7 system and seeks clarification on whether the error occurs at the beginning or during the run. The conversation highlights the importance of ensuring correct parameter settings in MCNPX for accurate simulation results.
saidhorion
Messages
2
Reaction score
0
hi every body i use MCNPX with a Z600 with 20Go of ram and 16 core to calculate the dep energy arrownd a source of cobalt with the tally F6, but when i put NPS=3E+9 an erreur apears that '' all entries must be integers''. i want help because NPS=2E+9 dosn't give a satisfaying results?
 
Engineering news on Phys.org
Hi,
i have the version: 2.7 of MCNPX, win7 64-bit.
 
So I suppose you don't have a fatal error on the beginning of the run but in progress ?
 
Hello everyone, I am currently working on a burnup calculation for a fuel assembly with repeated geometric structures using MCNP6. I have defined two materials (Material 1 and Material 2) which are actually the same material but located in different positions. However, after running the calculation with the BURN card, I am encountering an issue where all burnup information(power fraction(Initial input is 1,but output file is 0), burnup, mass, etc.) for Material 2 is zero, while Material 1...

Similar threads

Back
Top