Looking for MCNP tutorials for a beginner • Physics Forums
Jul 26, 2024 · The discussion reflects a range of experiences with MCNP, highlighting the need for tailored guidance for beginners. Individuals new to MCNP, researchers looking for simulation tools in …
MCNP TR Transform Card Question - Physics Forums
Nov 21, 2024 · Learn about translating coordinates in MCNP after applying transformations Investigate compatible cross-section tables for MCNP version 6.3 Explore troubleshooting techniques for …
Question about multiple runtimes on MPI for MCNP6
Nov 2, 2024 · The MCNP manual states that the command to use multiple CPUs is 'mpirun -np X', where X is the number of CPUs, followed by mcnp6.mpi (MCNP6 compiled for MPI) and input/output …
MCNP6: error -- "bad trouble in imcn" is usually ... - Physics Forums
Sep 4, 2023 · Hello, I'm new to MCNP deck-building, and I'm trying to acquire an X-ray energy spectrum using MCNP6, on Windows 10 environment. I'm running MCNP6 via MCNPX Visual Editor Version …
Understanding MCNP Tally F5 Output: Tips for Beginners
Aug 29, 2023 · This discussion focuses on interpreting the output of MCNP Tally F5, specifically regarding collided and uncollided photon flux. Beginners are guided to consider both types of flux for …
Help with F4 Fm4 dose calculation in MCNP simulation
Dec 3, 2024 · TL;DR Struggling with dose calculation in MCNP for brachytherapy using Ho-166. Using Fm -1 0 -5 -6 for photons results in zero, unlike F6/F8. Need help configuring F4:e,p with Fm4 to …
MCNP PTRAC card help - Physics Forums
May 12, 2025 · Hi all, I'm attempting to simulate a very specific setup in MCNP. I want to know the fraction of particles contributing to a surface tally that previously interacted (scattered) with a …
Beginner Seeking Help: MCNP6 Burnup Example (OECD-NEA Benchmark)
Aug 15, 2025 · 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 …
MCNP4 help: f4 tally in lattice - Physics Forums
May 22, 2022 · F4 is a flux tally, but instead of counting the neutrons going through it MCNP measures the total path length and divides it by the volume of the cell to calculate the flux. It's only the cookie …
MCNP Data Card Errors - Physics Forums
Feb 16, 2025 · Investigate MCNP material card formatting and common pitfalls Learn how to validate source energy distributions in MCNP Review best practices for avoiding unrecognized characters in …