last modified: 01-MAR-1965 | catalog | categories | new | search |

NEA-0157 SPM-046.

SPM-046, Reactor Kinetics by 1 Group Diffusion Calculation in R-Z Geometry

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1. NAME OR DESIGNATION OF PROGRAM:  SPM-046.
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2. COMPUTERS
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Program name Package id Status Status date
SPM-046 NEA-0157/01 Tested 01-MAR-1965

Machines used:

Package ID Orig. computer Test computer
NEA-0157/01 IBM 7094 IBM 7094
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3. NATURE OF PHYSICAL PROBLEM SOLVED

The time dependent diffusion equation in one neutron group is integrated for a reactor in r-z geometry. Transients due to axial movement of absorbers, or variation in coolant flow, can be studied. The model can be controlled to maintain a constant mean flux in the whole reactor, or to preserve axial flux distributions in three regions. Account is taken of the reactivity effects of temperature variation in fuel and moderator, and xenon poisoning.
Control is by axial movement of absorber situated anywhere in the reactor.
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4. METHOD OF SOLUTION

The calculation is performed step-wise in time. The diffusion calculation is integrated by a finite difference method with constant mesh intervals along each of the two directions.
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5. RESTRICTIONS ON THE COMPLEXITY OF THE PROBLEM

One to six delayed neutron groups. Five temperatures - fuel, canning, coolant, surface  of moderator and bulk moderator. A maximum of 20 x 30 mesh points can be accommodated. Radial and axial reflectors are simulated by extrapolation conditions. The thermal characteristics are assumed uniform with the exception of the coolant for which the radial distribution is estimated, assuming that the exit temperature is constant, and taking proper account of the thermal capacity of the moderator.
Overall control is a function of the temperature variation of the coolant at exit. The axial control is a function of the variation of the relation between the heating of the canning and the heating of the coolant.
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6. TYPICAL RUNNING TIME

Of the order of 0.03 seconds per step and per  mesh point. The time difference between steps should be of the order of a second for the study of fast transients, but may take values of the order of minutes when slowly varying phenomena are studied.
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7. UNUSUAL FEATURES OF THE PROGRAM

An initial flux distribution is given, allowing the simulation of a heterogeneous reactor.
The programme may be used with one radial point only which gives the possibility of extremely rapid analysis of purely axial problems.
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8. RELATED OR AUXILIARY PROGRAMS: RELATED AND AUXILIARY PROGRAMS
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9. STATUS
Package ID Status date Status
NEA-0157/01 01-MAR-1965 Tested at NEADB
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10. REFERENCES: - Note CEA no. 513.
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11. HARDWARE REQUIREMENTS: MACHINE REQUIREMENTS
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12. PROGRAMMING LANGUAGE(S) USED
Package ID Computer language
NEA-0157/01 FORTRAN+FAP
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13. OPERATING SYSTEM OR MONITOR UNDER WHICH PROGRAM IS EXECUTED:  FORTRAN II version 3.
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14. OTHER PROGRAMMING OR OPERATING INFORMATION OR RESTRICTIONS

ANY OTHER PROGRAMMING OR OPERATING INFORMATION OR RESTRICTIONS
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15. NAME AND ESTABLISHMENT OF AUTHOR

J.L. Cailly
                                   CEN Saclay
                                   France
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16. MATERIAL AVAILABLE
NEA-0157/01
File name File description Records
NEA0157_01.001 SOURCE & DATA 1836
NEA0157_01.002 OUTPUT 431
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17. CATEGORIES
  • F. Space - Time Kinetics, Coupled Neutronics - Hydrodynamics - Thermodynamics

Keywords: absorption, coolants, finite difference method, moderators, neutron diffusion equation, poisoning, r-z, temperature, time dependence, transients.