SCONE Guide

❮ ConfigurationTallies ❯

SCONE Verification Notice

SCONE is a research-oriented code with a smaller user base than MCNP, OpenMC, or SERPENT. Our examples are intended as educational guidance. For authoritative syntax, physics options, and nuclear data requirements, consult the official documentation.

Official SCONE documentationSCONE on GitHubIssues & discussions

Neutron Sources in SCONE

Defining Where Neutrons Begin Their Journey

Source Fundamentals

For the eigenvalue workflow used in your BEAVRS-style input, source treatment is driven by `pop`, `active`, and `inactive` plus fission-source convergence behavior.

plaintext
type eigenPhysicsPackage;

pop      100000;
active   300;
inactive 200;
XSdata ce;
dataType ce;

collisionOperator { neutronCE { type neutronCEstd; } }
transportOperator { type transportOperatorDT; }

! In k-eigenvalue mode, SCONE converges a fission source internally.
! There is no separate "source point" block like fixed-source examples
! in some other Monte Carlo codes.

This is the most important distinction: your production SCONE workflow is k-eigenvalue driven, so convergence control is the "source definition."

Advanced Source Definitions

In practice, you still monitor spatial behavior using mapped fission tallies and axial/radial diagnostics before trusting active-cycle statistics.

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! Practical source-convergence workflow
1) Start with conservative inactive cycles (e.g., 150-300 for full-core lattices).
2) Track cycle-wise k and spatial fission-rate maps.
3) Increase inactive cycles if source shape drifts across active cycles.
4) Only compare design variants after equivalent convergence behavior.

! Example active tally used in your deck:
activeTally {
  fissionRate {
    type collisionClerk;
    response (fission);
    fission { type macroResponse; MT -6; }
    map { type multiMap; maps (xax yax); ... }
  }
}

For SCONE tutorials tied to your input, this convergence workflow is far more useful than generic point-source syntax copied from other codes.

Best Practices

When defining sources, ensure the spatial distribution matches your geometry. Verify energy spectra against known physics (e.g., fission, fusion, or decay). Consider combining multiple sources for complex systems. Document your source parameters and their origins. Test source definitions with simple geometries before using in complex models.

Next Steps

With sources properly defined, you're ready to configure how neutrons travel through your system. The Transport section covers particle tracking methods, physics options, and variance reduction techniques that determine how your simulation runs.

Related Topics

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