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Depletion, activation, and decay with ORIGEN

Reactor and fuels analysis tools require accurate assessment of nuclide inventories using a general purpose predictive tools that can be applied to any system type. ORIGEN provides depletion, activation, and spent fuel source terms analysis capabilities are enabled through a family of modules are integrated with a variety of application tools including those from NEAMS, CASL, SCALE, and other programs. The nuclide tracking in ORIGEN is based on the principle of explicitly modeling all available nuclides and transitions in the current fundamental nuclear data for decay and neutron-induced transmutation and relies on fundamental cross section and decay data in ENDF/B VII. Cross section data for materials and reaction processes not available in ENDF/B-VII are obtained from the JEFF-3.0/A special purpose European activation library containing 774 materials and 23 reaction channels with 12,617 neutron-induced reactions below 20 MeV. Resonance cross section corrections in the resolved and unresolved range are performed using a continuous-energy treatment. All nuclear decay data, fission product yields, and gamma-ray emission data are developed from ENDF/B-VII.1 evaluations. Decay data include all ground and metastable state nuclides with half-lives greater than 1 millisecond. Using these data sources, ORIGEN currently tracks 174 actinides, 1149 fission products, and 974 activation products.

Two high-performance equation solvers are available: the hybrid linear chains and matrix exponential method and a new Chebyshev Rational Approximation Method (CRAM). Typical execution times are on the order of a few seconds for a multi-step solution, with each individual solution (step) taking approximately 10 milliseconds. ORIGEN also includes capabilities for continuous feed and removal by element. Output capabilities include isotopics (moles or grams), source spectra (alpha, beta, gamma, and neutron), activity (becquerels or curies), decay heat (total watts or gamma only), and radiological hazard factors (maximum permissible concentrations in air or water). These results can be displayed in the output file (.out extension) and/or archived in an ORIGEN binary results file (.f71 extension).

The ORIGEN reactor libraries distributed with SCALE include a set of pre-calculated ORIGEN libraries (with TRITON) for common fuel assembly types. These libraries may be used to rapidly assess spent fuel isotopics and source terms in these systems for arbitrary burnups and decay times.  Additional pre-computed libraries can be generated for additional reactor and fuel types:

•BWR 7×7, 8×8-1, 8×8-2, 9×9-2, 9×9-9, 10×10-9, 10×10-8, SVEA-64, SVEA-96, and SVEA-100
•PWR 14×14, 15×15, 16×16, 17×17, 18×18
•CANDU reactor (19-, 28-, and 37-element bundle designs)
•Magnox graphite reactor
•Advanced Gas-Cooled Reactor (AGR)
•VVER 440 and VVER 1000

•MOX BWR 7×7, 8×8-1, 8×8-2, 9×9-2, 9×9-9, 10×10-9, 10×10-8, SVEA-64, SVEA-96, 
and SVEA-100
•MOX PWR 14×14, 15×15, 16×16, 17×17, 18×18