NESC0467 HRG3
HRG-3, Slowing-Down Neutron Spectra Using P1 and B1 Approximation with Average Cross-Sections Calculation
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 AUTHORS, MATERIAL, CATEGORIES
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| Program name | Package id | Status | Status date |
|---|---|---|---|
| HRG-3 | NESC0467/01 | Tested | 01-JUN-1975 |
Machines used:
| Package ID | Orig. computer | Test computer |
|---|---|---|
| NESC0467/01 | IBM 370 series | IBM 370 series |
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3. DESCRIPTION OF PROBLEM OR FUNCTION
The code computes the slowing down spectrum over the energy range 10 MeV to 0.414 eV in either the B1 or P1 approximation, using 68 groups of neutrons with a constant group width of delta u = 0.25. The calculated flux and current spectra are used to reduce the original 68-group cross section data to average values over as many as 33 broad groups. Output is printed and may also be punched in formats for input to any of several spatial multigroup codes.
The code computes the slowing down spectrum over the energy range 10 MeV to 0.414 eV in either the B1 or P1 approximation, using 68 groups of neutrons with a constant group width of delta u = 0.25. The calculated flux and current spectra are used to reduce the original 68-group cross section data to average values over as many as 33 broad groups. Output is printed and may also be punched in formats for input to any of several spatial multigroup codes.
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4. METHOD OF SOLUTION
The 68 fine-group fluxes and currents are calculated by one sweep through the group structure, starting from a specified source distribution. The source may be selected from among the eight available on the data tape or may be input. The multigroup model uses a full downscattering matrix, with inelastic, (n,2n), and P0 and P1 components of elastic scattering explicitly included. Macroscopic fine-group parameters are constructed from input nuclide concentrations and microscopic parameters, available on the data tape for more than 200 individual nuclides. There is no restriction, other than availability, on the number of nuclides usable in a case. A special calculation is made in the resonance range for certain nuclides, using an adaptation of the Adler, Hinman, and Nordheim method to obtain an intermediate resonance approximation for both the absorber nuclide and an admixed moderator. The resonance contribution is allocated to the fine-groups in a consistent manner providing self-shielding in both space and energy. Additional self- shielding factors may be read in for any nuclide, if desired. The fine-group fluxes and currents are used as weighting functions in averaging macroscopic and microscopic parameters over the specific option, the neutron age in an infinite medium may be calculated by the moments method.
The 68 fine-group fluxes and currents are calculated by one sweep through the group structure, starting from a specified source distribution. The source may be selected from among the eight available on the data tape or may be input. The multigroup model uses a full downscattering matrix, with inelastic, (n,2n), and P0 and P1 components of elastic scattering explicitly included. Macroscopic fine-group parameters are constructed from input nuclide concentrations and microscopic parameters, available on the data tape for more than 200 individual nuclides. There is no restriction, other than availability, on the number of nuclides usable in a case. A special calculation is made in the resonance range for certain nuclides, using an adaptation of the Adler, Hinman, and Nordheim method to obtain an intermediate resonance approximation for both the absorber nuclide and an admixed moderator. The resonance contribution is allocated to the fine-groups in a consistent manner providing self-shielding in both space and energy. Additional self- shielding factors may be read in for any nuclide, if desired. The fine-group fluxes and currents are used as weighting functions in averaging macroscopic and microscopic parameters over the specific option, the neutron age in an infinite medium may be calculated by the moments method.
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NESC0467/01
| File name | File description | Records |
|---|---|---|
| NESC0467_01.001 | INFORMATION | 1 |
| NESC0467_01.002 | BINHDT JCL | 12 |
| NESC0467_01.003 | BINHDT SOURCE PROGRAM (F4) | 286 |
| NESC0467_01.004 | BINHDT SAMPLE PROBLEM DATA | 2 |
| NESC0467_01.005 | BINHDT BCD LIBRARY READ FROM UNIT 11 | 17635 |
| NESC0467_01.006 | BINHDT BCD LIBRARY READ FROM UNIT 13 | 19530 |
| NESC0467_01.007 | BINHDT SAMPLE PROBLEM PRINTED OUTPUT | 1080 |
| NESC0467_01.008 | NUTAPE SOURCE PROGRAM (F4) | 858 |
| NESC0467_01.009 | HRG-3 JCL | 8 |
| NESC0467_01.010 | HRG-3 SOURCE PROGRAM (F4) | 2116 |
| NESC0467_01.011 | HRG-3 SAMPLE PROBLEM DATA | 21 |
| NESC0467_01.012 | HRG-3 SAMPLE PROBLEM PRINTED OUTPUT | 597 |
Keywords: cross sections, elastic scattering, group constants, inelastic scattering, slowing-down.