5.1. ORIGEN: Neutron activation, transmutation, fission product generation, & radiation source term calculation

ORIGEN (Oak Ridge Isotope Generation code) calculates time-dependent concentrations, activities, and radiation source terms for a large number of isotopes simultaneously generated or depleted by neutron transmutation, fission, and radioactive decay. ORIGEN is used internally within SCALE’s TRITON and Polaris sequences to perform depletion and decay. As a stand-alone SCALE module, ORIGEN provides additional unique capabilities to (1) simulate continuous nuclide feed and chemical removal, which can be used to model reprocessing or liquid fuel systems, and (2) generate alpha, beta, neutron and gamma decay emission spectra. A standard decay library is provided to perform decay calculations. For neutron activation and fuel depletion problems, neutron spectrum-dependent ORIGEN libraries are required and may be created from (1) user-defined spectrum and self-shielded cross sections using the COUPLE module or (2) interpolation of existing ORIGEN reactor libraries (precalculated by TRITON) using the Automated Rapid Processing (ARP) module. Post-processing using the OPUS module allows calculated isotopics and spectra to be sorted, ranked, and converted to other units.

• 5.1.1. Acknowledgements
• 5.1.2. Version Information
  • 5.1.2.1. Version 6.3 (2021)
  • 5.1.2.2. Version 6.2 (2016)
  • 5.1.2.3. Version 6.1 (2011)
    • 5.1.2.3.1. ORIGEN
    • 5.1.2.3.2. COUPLE
    • 5.1.2.3.3. ARP
    • 5.1.2.3.4. OPUS
• 5.1.3. Method of solution
  • 5.1.3.1. Methodology
    • 5.1.3.1.1. Generation of Problem-dependent Transition Matrix
    • 5.1.3.1.2. Solution of the Depletion/Decay Equations
      • 5.1.3.1.2.1. MATREX
        • Solution for Long-Lived Nuclides
        • Solution for Short-Lived Nuclides
        • Solution of the Nonhomogeneous Equation
      • 5.1.3.1.2.2. CRAM
    • 5.1.3.1.3. Power Calculation
    • 5.1.3.1.4. Decay Emission Sources Calculation
      • 5.1.3.1.4.1. Neutron Sources
      • 5.1.3.1.4.2. Alpha Sources
      • 5.1.3.1.4.3. Beta Sources
      • 5.1.3.1.4.4. Photon Sources
    • 5.1.3.1.5. Library Interpolation
      • 5.1.3.1.5.1. Lagrangian Interpolation
      • 5.1.3.1.5.2. Cubic Spline Interpolation
• 5.1.4. ORIGEN Family of Modules
• 5.1.5. ORIGEN
  • 5.1.5.1. Key Features
    • 5.1.5.1.1. Nuclide Specification and ORIGEN Sub-libraries
    • 5.1.5.1.2. Physical Units in Calculations
    • 5.1.5.1.3. Saving Results
  • 5.1.5.2. Input Description
    • 5.1.5.2.1. Calculation Case (case)
    • 5.1.5.2.2. Transition Matrix Specification (lib)
    • 5.1.5.2.3. Material Specification (mat)
    • 5.1.5.2.4. Operating History (power, flux, time)
    • 5.1.5.2.5. Printing Options (print)
      • 5.1.5.2.5.1. Inventories by nuclide and element
      • 5.1.5.2.5.2. Radiological Emissions (alpha, beta, gamma, neutron)
        • Neutron print options
        • Gamma print options
        • Alpha print options
        • Beta print options
    • 5.1.5.2.6. Saving Results (save)
    • 5.1.5.2.7. Decay Emission Calculations (alpha, beta, gamma, neutron)
      • 5.1.5.2.7.1. Neutron source calculation
      • 5.1.5.2.7.2. Gamma source calculation
      • 5.1.5.2.7.3. Alpha and beta source calculation
    • 5.1.5.2.8. Processing Options (processing)
    • 5.1.5.2.9. Bounds Block
    • 5.1.5.2.10. Solver Block
    • 5.1.5.2.11. Options Block
    • 5.1.5.2.12. Library Building Block build_lib
      • 5.1.5.2.12.1. Nuclide Block build_lib/nuclide
      • 5.1.5.2.12.2. Decay Data Block build_lib/decay
        • Updating specific decay coefficients
        • Notes on Spontaneous Fission
      • 5.1.5.2.12.3. Neutron Data Block build_lib/neutron
      • 5.1.5.2.12.4. Data Cutoffs
    • 5.1.5.2.13. Sensitivity Calculation Block sens
      • 5.1.5.2.13.1. Sensitivity output tables
      • 5.1.5.2.13.2. Sensitivity output files
• 5.1.6. ARP
  • 5.1.6.1. Input Description
  • 5.1.6.2. ARPDATA.TXT listing file
• 5.1.7. OPUS
  • 5.1.7.1. Key Features
  • 5.1.7.2. Input Description
    • 5.1.7.2.1. Nuclide/element plots
    • 5.1.7.2.2. Spectra plots
    • 5.1.7.2.3. Selection of plotting data
  • 5.1.7.3. Plot File Formats
• 5.1.8. Examples
  • 5.1.8.1. Decay of ²³⁸U
  • 5.1.8.2. ²⁵²Cf neutron Emission Spectrum
  • 5.1.8.3. Simple Fuel Irradiation Plus Decay
  • 5.1.8.4. Three Cycles of Irradiation Plus Decay
  • 5.1.8.5. Load Isotopics from an f71 File
  • 5.1.8.6. Continuous Feed and Removal
  • 5.1.8.7. Calculate Fuel \(\left(\alpha,n \right)\) Emissions in a Glass Matrix
  • 5.1.8.8. Create an ORIGEN Decay Library from a Decay Resource
  • 5.1.8.9. Create an ORIGEN Reaction Library
  • 5.1.8.10. Create an ORIGEN Activation Library
  • 5.1.8.11. Create an ORIGEN Library with User-Supplied Cross Sections
  • 5.1.8.12. Printing library cross-section values
    • 5.1.8.12.1. Print the Cross Section Values on an ORIGEN Library with OBIWAN
    • 5.1.8.12.2. Print the Cross Section Values on an ORIGEN Library in ORIGEN
  • 5.1.8.13. Ranking Contribution to Toxicity
  • 5.1.8.14. Spectrum plots with OPUS
    • 5.1.8.14.1. Photon Spectrum Plot
    • 5.1.8.14.2. Neutron Spectrum Plot
  • 5.1.8.15. Isotopic Weight Percentages for Uranium and Plutonium During Decay
  • 5.1.8.16. User-Specified Response Function in OPUS

References

ORIGEN-BA67

S. J. Ball and R. K. Adams. MATEXP: A General Purpose Digital Computer Program for Solving Ordinary Differential Equations by the Matrix Exponential Method. Technical Report ORNL/TM-1933, Union Carbide Corporation (Nuclear Division), Oak Ridge National Laboratory, 8 1967.

ORIGEN-Bat10

H. Bateman. The Solution of Differential Equations Occurring in the Theory of Radioactive Transformations. Proc. Cambridge Phil. Soc., 15:423, 1910.

ORIGEN-Bel73

M. J. Bell. ORIGEN B-The ORNL Isotope Generation and Depletion Code . Technical Report ORNL-4628 (CCC-217), Union Carbide Corporation (Nuclear Division), Oak Ridge National Laboratory, 5 1973.

ORIGEN-DJPB86

D. J. Pellarin, W. L. Matney and N. E. Bibler. :math:`\left (\alpha ,n \right )` neutron emission from dwpf glass. Technical Report DPST-86-212, Savannah River Laboratory, 1 1986. URL: https://www.osti.gov/servlets/purl/780500-Z8ooL7/native/.

ORIGEN-Gau03

I. C. Gauld. MOX Cross-Section Libraries for ORIGEN-ARP. Technical Report ORNL/TM-2003/2, UT-Battelle, LLC, Oak Ridge National Laboratory, Oak Ridge, TN (USA), 7 2003.

ORIGEN-IA11

A. E. Isotalo and P. A. Aarnio. Comparison of depletion algorithms for large systems of nuclides. Annals of Nuclear Energy, 38(2):261 – 268, 2011. URL: http://www.sciencedirect.com/science/article/pii/S0306454910003889, doi:https://doi.org/10.1016/j.anucene.2010.10.019.

ORIGEN-LL67

L. Lapidus and R. Luus. Optimal Control of Engineering Processes, pages 45–49. Blaisdell Publishing Co., Waltham, MA, 1967.

ORIGEN-LHBP98

L. C. Leal, O. W. Hermann, S. M. Bowman, and C. V. Parks. ARP: Automatic Rapid Process for the Generation of Problem-Dependent SAS2H/ORIGEN-S Cross-Section Libraries. Technical Report ORNL/TM-13584, Lockheed Martin Energy Research Corporation, Oak Ridge National Laboratory, 4 1998.

ORIGEN-ORN11

ORNL. SCALE: A Comprehensive Modeling and Simulation Suite for Nuclear Safety Analysis and Design". Technical Report ORNL/TM-2005/39, Version 6.1, Oak Ridge National Laboratory, Oak Ridge, TN, 6 2011.

ORIGEN-Pus11

Maria Pusa. Rational Approximations to the Matrix Exponential in Burnup Calculations. Nuclear Science and Engineering, 169(2):155–167, 2011. URL: https://doi.org/10.13182/NSE10-81, doi:10.13182/NSE10-81.

ORIGEN-Pus13

Maria Pusa. Numerical Methods for Nuclear Fuel Burnup Calculations. PhD thesis, Aalto University, 2013.

ORIGEN-PL10

Maria Pusa and Jaakko Leppänen. Computing the Matrix Exponential in Burnup Calculations. Nuclear Science and Engineering, 164(2):140–150, 2010. doi:10.13182/NSE09-14.

ORIGEN-Sho00

E. F. Shores. Data Updates for the SOURCES-4A Computer Code. Technical Report LA-UR-00-5016, Los Alamos National Laboratory, 10 2000. SOURCES-4C available from the Radiation Safety Information Computational Center (RSICC) as code package C00661.

ORIGEN-Von62

D. R. Vondy. Development of a General Method of Explicit Solution to the Nuclide Chain Equations for Digital Machine Calculations. Technical Report ORNL/TM-361, Union Carbide Corporation (Nuclear Division), Oak Ridge National Laboratory, 10 1962.

ORIGEN-Wil86

Mark L. Williams. Perturbation theory for nuclear reactor analysis. CRC Press, Inc., 1986.

ORIGEN-WPC+99

W. B. Wilson, R. T. Perry, W. S. Charlton, T. A. Parish, G. P. Estes, T. H. Brown, E. D. Arthur, M. Bozoian, T. R. England, D. G. Madland, and J. E. Stewart. SOURCES 4A: Code for Calculating :math:`\left (\alpha ,n\right )`, Spontaneous Fission, and Delayed Neutron Sources and Spectra. Technical Report LA-13639-MS, Los Alamos National Laboratory, 9 1999.

ORIGEN-WPS+83

W. B. Wilson, R. T. Pery, J. E. Stewart, T. R. England, D. G. Madland, and E. D. Arthur. Development of the SOURCES Code and Data Library for the Calculation of Neutron Sources and Spectra from :math:`\left ( \alpha ,n \right )` Reactions, Spontaneous Fission, and :math:`beta^-` Delayed Neutrons. Technical Report LA-9841-PR, Los Alamos National Laboratory, Los Alamos, NM, 1983.

ORIGEN-WA02

George Wolberg and Itzik Alfy. An energy-minimization framework for monotonic cubic spline interpolation. Journal of Computational and Applied Mathematics, 143:145–188, 6 2002.

ORIGEN-REFed68

R. E. Funderlic (ed.). The programmer’s handbook-a compendium of numerical analysis utility programs. Technical Report AEC Research and Development Report K-1729, Oak Ridge National Laboratory, Oak Ridge, TN (USA), 2 1968.

5.1.9. Appendices

• 5.1.9.1. PRISM
• 5.1.9.2. ARPLIB
• 5.1.9.3. XSECLIST
• 5.1.9.4. COUPLE
  • 5.1.9.4.1. Key Features
    • 5.1.9.4.1.1. AMPX multi-group libraries
    • 5.1.9.4.1.2. Nuclide Specification
    • 5.1.9.4.1.3. Adding new transitions and user-defined transitions
    • 5.1.9.4.1.4. Unit numbers and Aliases
  • 5.1.9.4.2. Input Description
    • 5.1.9.4.2.1. Block1: titles, unit numbers, and case controls.
    • 5.1.9.4.2.2. Block2: nuclides with fission yields and weighting flux spectrum
    • 5.1.9.4.2.3. Block3: array dimensions for decay library creation
    • 5.1.9.4.2.4. Block6: number of user-defined transition coefficients
    • 5.1.9.4.2.5. Block8: user-defined transition coefficients