Computer Programs

NAME OR DESIGNATION OF PROGRAM, COMPUTER, DESCRIPTION OF PROBLEM OR FUNCTION, METHOD OF SOLUTION, RESTRICTIONS ON THE COMPLEXITY OF THE PROBLEM, TYPICAL RUNNING TIME, UNUSUAL FEATURES OF THE PROGRAM, RELATED AND AUXILIARY PROGRAMS, STATUS, REFERENCES, MACHINE REQUIREMENTS, LANGUAGE, OPERATING SYSTEM UNDER WHICH PROGRAM IS EXECUTED, OTHER PROGRAMMING OR OPERATING INFORMATION OR RESTRICTIONS, NAME AND ESTABLISHMENT OF AUTHOR, MATERIAL, CATEGORIES

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To submit a request, click below on the link of the version you wish to order. Rules for end-users are
available here.

Program name | Package id | Status | Status date |
---|---|---|---|

ORIGEN-JR | NEA-0622/01 | Tested | 11-JUN-1982 |

Machines used:

Package ID | Orig. computer | Test computer |
---|---|---|

NEA-0622/01 | IBM 3033 | IBM 3033 |

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3. DESCRIPTION OF PROBLEM OR FUNCTION

ORIGEN-JR, which has been developed from ORIGEN, calculates radiation sources and analyses nuclide transmutations. The calculation of neutron source spectra has been revised extensively. In addition to the spectrum indices adopted in ORIGEN, one-group cross sections for each reaction can be used to treat precisely the burnup conditions in a reactor core. Neutron and gamma-ray source data are generated in the same format as in the shielding codes QAD-PS, ANISN and DOT.

ORIGEN-JR, which has been developed from ORIGEN, calculates radiation sources and analyses nuclide transmutations. The calculation of neutron source spectra has been revised extensively. In addition to the spectrum indices adopted in ORIGEN, one-group cross sections for each reaction can be used to treat precisely the burnup conditions in a reactor core. Neutron and gamma-ray source data are generated in the same format as in the shielding codes QAD-PS, ANISN and DOT.

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4. METHOD OF SOLUTION

The code solves the equations of radioactive nuclide transmutation and calculates radiation sources. Energy spectra of neutron sources in spontaneous fission and (alpha, n) reactions of eight nuclides (9Be, 10B, 11B, 13C, 14N, 170, 180 and 19F) are available. Complex decay and transmutation schemes and one-group reaction cross sections can be treated.

A Matrix Exponential method is used to solve the resultant large system of coupled, linear, first-order, ordinary differential equations with constant coefficients.

The code solves the equations of radioactive nuclide transmutation and calculates radiation sources. Energy spectra of neutron sources in spontaneous fission and (alpha, n) reactions of eight nuclides (9Be, 10B, 11B, 13C, 14N, 170, 180 and 19F) are available. Complex decay and transmutation schemes and one-group reaction cross sections can be treated.

A Matrix Exponential method is used to solve the resultant large system of coupled, linear, first-order, ordinary differential equations with constant coefficients.

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5. RESTRICTIONS ON THE COMPLEXITY OF THE PROBLEM

The code will handle 850 nuclides of which no more than 500 are described as light elements, no more than 120 are actinides and their decay daughters, and no more than 500 are fission products.

Fission products are produced by fission rates of five different actinides. In addition, there may be no more than 3000 non-zero elements in the nuclear transmutation matrix. Numerical inaccuracies due to too coarse a time spacing are indicated by a warning message.

Neutron spectra due to (alpha, n) reaction can be calculated by no more than 61 actinides and 8 light nuclides.

The code will handle 850 nuclides of which no more than 500 are described as light elements, no more than 120 are actinides and their decay daughters, and no more than 500 are fission products.

Fission products are produced by fission rates of five different actinides. In addition, there may be no more than 3000 non-zero elements in the nuclear transmutation matrix. Numerical inaccuracies due to too coarse a time spacing are indicated by a warning message.

Neutron spectra due to (alpha, n) reaction can be calculated by no more than 61 actinides and 8 light nuclides.

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10. REFERENCES

- M.J. Bell:

ORIGEN - The ORNL Isotope Generation and Depletion Code

ORNL-4628 (1973).

- D.L. Lessor, R.E. Schenter:

Neutron Spectra from (Alpha, n) Reactions in Plutonium Compounds

Calculated from Hauser-Feshbach Reaction Theory

BNWL-B-109 (1971).

- S. Igarasi:

Program ELIESE-3 - Program for Calculation of the Nuclear Cross

Sections by Using Local and Non-Local Optical Models and

Statistical Models

JAERI-1224 (1972).

- K. Koyama, et al.:

RADHEAT-V3 - A Code System for Generating Coupled Neutron and

Gamma-Ray Group Constants and Analyzing Radiation Transport

JAERI-M 7155 (1977).

- M.J. Bell:

ORIGEN - The ORNL Isotope Generation and Depletion Code

ORNL-4628 (1973).

- D.L. Lessor, R.E. Schenter:

Neutron Spectra from (Alpha, n) Reactions in Plutonium Compounds

Calculated from Hauser-Feshbach Reaction Theory

BNWL-B-109 (1971).

- S. Igarasi:

Program ELIESE-3 - Program for Calculation of the Nuclear Cross

Sections by Using Local and Non-Local Optical Models and

Statistical Models

JAERI-1224 (1972).

- K. Koyama, et al.:

RADHEAT-V3 - A Code System for Generating Coupled Neutron and

Gamma-Ray Group Constants and Analyzing Radiation Transport

JAERI-M 7155 (1977).

NEA-0622/01, included references:

- K. Koyama, N. Yamano, S. Miyasaka:A computer code for calculating radiation sources and analizing

nuclide transmutations, JAERI-M 8229 (May 1979)

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NEA-0622/01

File name | File description | Records |
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NEA0622_01.003 | ORIGEN-JR INFORMATION FILE | 152 |

NEA0622_01.004 | ORIGEN-JR, SOURCE CARD IMAGES (FORTRAN-4) | 3856 |

NEA0622_01.006 | ORIGEN-JR, INPUT DATA FOR SAMPLE CASE | 39 |

NEA0622_01.007 | NUCLIDE LIB DATA (LIGHT ELEMENTS) | 1265 |

NEA0622_01.008 | NUCLIDE LIB DATA (HEAVY ELEMENTS) | 505 |

NEA0622_01.009 | FISSION PRODUCTS (NUCLIDES) | 2305 |

NEA0622_01.010 | PHOTON LIB DATA (LIGHT ELEMENTS) | 253 |

NEA0622_01.011 | PHOTON LIB DATA (HEAVY ELEMENTS) | 202 |

NEA0622_01.012 | FISSION PRODUCTS (PHOTONS) | 461 |

NEA0622_01.013 | (ALPHA,N) LIB DATA | 7409 |

NEA0622_01.014 | (ALPHA,N) LIB DATA PREPARATION PROGRAM | 23 |

NEA0622_01.015 | STOPPING POWER LIB DATA | 232 |

NEA0622_01.016 | STOPPING POWER LIB DATA PREPARATION PROGRAM | 14 |

NEA0622_01.017 | ORIGEN-JR, OUTPUT OF SAMPLE CASE | 4093 |

NEA0622_01.018 | ORIGEN-JR, JCL TO RUN SAMPLE CASE | 129 |

Keywords: burnup, cross sections, decay, fission products, gamma radiation, neutron spectra, optical models, radioactivity.