Recent content by HEU_LL

Hello everyone, I’m trying to use the JEFF-3.3 Nuclear Data Librar with MCNP6 . Could someone please provide detailed guidance or a step-by-step tutorial on how to properly configure MCNP6 to work with JEFF-3.3? Any advice, configuration files, or references would be greatly appreciated. Thank...

I'm very sorry to disturb everyone. The problem has been solved. In fact, it was due to carelessness that I wrote the wrong number of the materials that needed to be burned. I'm new to this forum and still don't know how to delete this help-seeking post.

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...

Thank you very much for your enthusiasm and patient guidance. I deeply admire your excellent professional knowledge. With the two approaches you mentioned above, I have successfully resolved the issue. Once again, I sincerely appreciate your help. Wishing you all the best in your endeavors.

Thank you very much for your patient guidance. Actually, I added Am241 and Am243 to the initial composition of the burnup materials and set them to a minute quantity of 1.0E-36, but the code failed to run and reported the following errors: “ fatal error. cross-section tables missing for zaid =...

Thank you so much for your enthusiastic reply, which has been very encouraging. Based on your suggestions, I carefully studied the usage rules of the BURN card, but I did not find data for Am241 and Am243 in the output results. How should this issue be addressed?

Hi everyone, I'm a complete beginner with MCNP and trying to learn how to perform burnup calculations. Right now, I'm feeling a bit lost and not sure where to start. I found the OECD-NEA Burnup Credit Calculational Criticality Benchmark (Phase I-B) and was wondering if anyone has worked through...