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Classes

Table of Contents

General Information

  • Before registering, you must acquire your own version of MCNP6 (see How to Get the MCNP Code).
  • All students who attend classes held in Los Alamos, NM must arrive with the MCNP code installed on their laptop for use in the classroom. It is strongly recommended that the installation DVDs also be brought (and a way to read them) in case the code needs to be reinstalled on the student's laptop during the class. In addition to the installation instructions provided with MCNP6.2, a YouTube tutorial for installing MCNP6.2 on Microsoft Windows is available. Similarly, in addition to the installation instructions provided with MCNP6.3, a YouTube tutorial for installing MCNP6.3 on Microsoft Windows is available.
  • Classes are open to all, but non-US-citizens must apply well in advance. Firm registration deadlines are listed in the schedule.
  • Registration for US-citizens will usually close two weeks prior to the start of a class.
  • The "go / no-go" determination date listed for each course is the date on which the enrollment will be considered to determine whether the class will be held or cancelled. This is also the date after which withdrawal from a class for reasons other than class cancellation becomes nonrefundable. Those hoping to attend a class should register well ahead of this date to ensure enrollment numbers are accurate.
  • For all questions on MCNP classes, please contact the MCNP Registrar.

Upcoming Classes

When Where What How
June 2–6, 2025
  Add to Calendar		 Online MCNP6 for Nuclear Safeguards Practitioners
Non-US citizens must register by March 14, 2025.
Go / no-go determination date is May 05, 2025.		 $600
June 9–13, 2025
  Add to Calendar		 Online Introduction to MCNP6
Non-US citizens must register by March 21, 2025.
Go / no-go determination date is May 12, 2025.		 Full
September 8–12, 2025
  Add to Calendar		 Los Alamos, NM Using NJOY to Create MCNP ACE Files and Visualize Nuclear Data
Non-US citizens must register by June 20, 2025.
Go / no-go determination date is August 11, 2025.		 $1800
October 6–10, 2025
  Add to Calendar		 Los Alamos, NM Intermediate MCNP6
Non-US citizens must register by July 18, 2025.
Go / no-go determination date is September 08, 2025.		 $1800
October 27–31, 2025
  Add to Calendar		 Online Introduction to MCNP6
Non-US citizens must register by August 08, 2025.
Go / no-go determination date is September 29, 2025.		 $600
December 1–5, 2025
  Add to Calendar		 Los Alamos, NM Variance Reduction with MCNP6
Non-US citizens must register by September 12, 2025.
Go / no-go determination date is November 03, 2025.		 $1800

  Add All Listed MCNP Classes to Calendar

Class Topics

Introduction to MCNP6

  • New features in MCNP6.
  • Input construction and basic geometry.
  • Plotting of geometry.
  • Source definitions.
  • Tallies and plotting.
  • Neutron data and photon/electron physics.
  • Statistical analysis.
  • Criticality.
  • Variance reduction.

    • Intermediate MCNP6

  • New features in MCNP6.
  • Very brief refresher of basic class.
  • Advanced geometry: universes, lattices, repeated structures.
  • Advanced source definitions: repeated sources, surface source read/write.
  • Advanced tallies: repeated tallies, perturbation tallies, pulse-height tallies.
  • Advanced variance reduction: dxtran, weight windows.
  • MCNPTools and more advanced utilities.

    • Criticality Calculations with MCNP6

  • Criticality calculations using Monte Carlo methods.
  • Geometry, including some advanced features.
  • Tallies & mesh tallies.
  • Cross-section data.
  • Statistical analysis.
  • Assessing convergence of Keff and the source distribution.
  • Interactive running and plotting with MCNP.

    • MCNP6 for Nuclear Safeguards Practitioners

  • MCNP basics overview and refresher.
  • Advanced geometry (universes, lattices, repeated structures).
  • Nuclear data libraries and material definitions.
  • Fixed source definitions.
  • Advanced tallies.
  • Non-destructive assay (NDA) system optimization exercise.
  • Other advanced topics, such as MCNPTools, may be included time permitting and depending on the interests of the course participants.

    • Practical MCNP for the Health Physicist, Radiological Engineer, and Medical Physicist

  • The Monte Carlo method and the importance of statistics.
  • Standard MCNP input deck layout.
  • Geometry: surfaces and cells.
  • Physics, material, source definition cards.
  • Variance Reduction: importance values.
  • Tallies: fluence, dose, and exposure.

    • Unstructured Mesh with Attila4MC

  • Introduction to the MCNP6 unstructured mesh / hybrid geometry capability.
  • Overview of Attila4MC and CAD modeling.
  • Introduction to SpaceClaim.
  • Creating models with SpaceClaim.
  • CAD cleanup / defeaturing with SpaceClaim.
  • Problem setup with the Attila4MC GUI.
  • Student exercise: creating an input from scratch.
  • Post-processing visualization.
  • Using Attila4MC automated importance generation for variance reduction.
  • Introduction to deterministic transport & deterministic weight windows.
  • FW-CADIS and CADIS with Attila4MC using CAD models.
  • Introduction to multiple unstructured meshes and CSG hybrid geometry in MCNP6.
  • MCNP6 unstructured mesh pre- and post-processors.
  • Introduction to MCNP activation and shutdown dose rate calculations using Attila4MC.
  • Student exercises.
  • Open lab.

    • Using NJOY to Create MCNP ACE Files and Visualize Nuclear Data

  • Introduction to ENDF.
  • History of NJOY.
  • How to obtain the code package and update the source code to create a new executable.
  • Create input decks and examine the associated output from the basic components of NJOY required to produce MCNP continuous energy neutron ACE files and MCNP thermal scattering ACE files.
  • Create input decks for NJOY's PLOTR & VIEWR modules to visualize cross sections and their ratios, scattering angular distributions, & secondary spectra.
  • Run MCNP with newly processed data.
  • Create multigroup data files using NJOYs GROUPR module.
  • Convert incident neutron nuclear data files in the ENDF format into MCNP continuous energy (.c) ACE files.
  • Plotting nuclear data using NJOY.
  • Convert ENDF formatted thermal scattering law files into MCNP (.t) ACE files.
  • Utilize newly processed data in MCNP.
  • Processing nuclear data into multigroup data files.

    • Variance Reduction with MCNP6

  • Brief review of particle transport, tallying, and statistics.
  • Sampling, biasing, and statistical convergence.
  • Understanding particle state with event logs.
  • Implicit capture & weight rouletting review.
  • Geometry-based importance splitting & rouletting review.
  • Source biasing.
  • Weight windows, and ways to generate them.
  • Point detectors and DXTRAN.
  • Forced collisions.
  • Exponential transform.