NESC0491 MOD5
MOD-5, Time-Dependent MultiGroup Slowing-Down Neutron Spectra and Keff Calculation, Green Function Method
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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| Program name | Package id | Status | Status date |
|---|---|---|---|
| MOD-5 | NESC0491/01 | Arrived | 01-FEB-2002 |
Machines used:
| Package ID | Orig. computer | Test computer |
|---|---|---|
| NESC0491/01 | IBM 360 series |
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3. DESCRIPTION OF PROBLEM OR FUNCTION
MOD5 calculates the time- and
energy-dependent evolution of the neutron density in homogeneous
media following initiation of a) a monoenergetic source
distributed over a finite time interval, or b) a source of
arbitrary spectrum with a delta-function distribution in time.
Effectively the code produces Green's function solutions to the
slowing-down equation in discrete numerical form. Leakage is
treated in the diffusion approximation. The program a) calculates
spectra and energy moments at selected times following the burst
of source neutrons, b) evaluates the time-dependent neutron
density and slowing-down density at selected energies and computes
moments of these densities, c) calculates time-dependent
distributions of capture, leakage and first fission, and moments
of these distributions, d) calculates steady-state central core
neutron flux and leakage flux in detail and in group-averaged
form, and e) calculates parameters such as keff.
MOD5 calculates the time- and
energy-dependent evolution of the neutron density in homogeneous
media following initiation of a) a monoenergetic source
distributed over a finite time interval, or b) a source of
arbitrary spectrum with a delta-function distribution in time.
Effectively the code produces Green's function solutions to the
slowing-down equation in discrete numerical form. Leakage is
treated in the diffusion approximation. The program a) calculates
spectra and energy moments at selected times following the burst
of source neutrons, b) evaluates the time-dependent neutron
density and slowing-down density at selected energies and computes
moments of these densities, c) calculates time-dependent
distributions of capture, leakage and first fission, and moments
of these distributions, d) calculates steady-state central core
neutron flux and leakage flux in detail and in group-averaged
form, and e) calculates parameters such as keff.
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4. METHOD OF SOLUTION
The code is based on a discrete Markov
slowing-down model developed by the author. The energy range of
interest in the slowing-down region is divided into an arbitrary
number of states and a Markov matrix is constructed which defines
the probabilities for transition from one state to another in some
finite time interval. The neutron density (defined as a vector
over the state structure) is evolved in time by repetitive
multiplication of the density vector into the transition matrix.
Currently cross sections are derived from the 26-group Russian
set. Capture, fission, and leakage events are accounted for in
three absorbing states.
The code is based on a discrete Markov
slowing-down model developed by the author. The energy range of
interest in the slowing-down region is divided into an arbitrary
number of states and a Markov matrix is constructed which defines
the probabilities for transition from one state to another in some
finite time interval. The neutron density (defined as a vector
over the state structure) is evolved in time by repetitive
multiplication of the density vector into the transition matrix.
Currently cross sections are derived from the 26-group Russian
set. Capture, fission, and leakage events are accounted for in
three absorbing states.
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5. RESTRICTIONS ON THE COMPLEXITY OF THE PROBLEM
Maxima of -
5 isotopes
71 real energy states
201 virtual energy states
26 broad groups for input cross sections
400 time-steps
21 time moments
10 groups from which inelastic scattering can take place
The program has been modified to handle up to 701 virtual energy
states and 201 real states.
Maxima of -
5 isotopes
71 real energy states
201 virtual energy states
26 broad groups for input cross sections
400 time-steps
21 time moments
10 groups from which inelastic scattering can take place
The program has been modified to handle up to 701 virtual energy
states and 201 real states.
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7. UNUSUAL FEATURES OF THE PROGRAM
(a) Elastic scattering transition probabilities are calculated with a new stochastic algorithm, reference 1.
(b) The state structure can be calculated by the code to provide optimal (most accurate) treatment of elastic scattering.
(c) Calculation times may be greatly reduced by allowing the transition matrix to follow the neutron distribution.
This traveling array technique may also allow a much finer state structure for a given core capacity.
(d) A unique transition matrix generating technique provides transition matrices that are consistent, accurate, and stable regardless of energy range or time-step width.
(a) Elastic scattering transition probabilities are calculated with a new stochastic algorithm, reference 1.
(b) The state structure can be calculated by the code to provide optimal (most accurate) treatment of elastic scattering.
(c) Calculation times may be greatly reduced by allowing the transition matrix to follow the neutron distribution.
This traveling array technique may also allow a much finer state structure for a given core capacity.
(d) A unique transition matrix generating technique provides transition matrices that are consistent, accurate, and stable regardless of energy range or time-step width.
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10. REFERENCES
T. J. Williamson and R. W. Albrecht, A Direct
Stochastic Model for Neutron Moderation, Nuclear Science and
Engineering, Vol. 37, pp. 41-58, 1969.
T. J. Williamson and R. W. Albrecht, Calculations of
Neutron Time-Energy Distributions in Heavy Moderators, Nuclear
Science and Engineering, Vol. 42, pp. 89-111, 1970.
L. P. Abagyan, et al., Group Constants for Nuclear
Reactor Calculations, Consultants Bureau, 1964.
T. J. Williamson, MOD5 User's Manual, NPS-61WN71061A,
1971.
T. J. Williamson and R. W. Albrecht, A Direct
Stochastic Model for Neutron Moderation, Nuclear Science and
Engineering, Vol. 37, pp. 41-58, 1969.
T. J. Williamson and R. W. Albrecht, Calculations of
Neutron Time-Energy Distributions in Heavy Moderators, Nuclear
Science and Engineering, Vol. 42, pp. 89-111, 1970.
L. P. Abagyan, et al., Group Constants for Nuclear
Reactor Calculations, Consultants Bureau, 1964.
T. J. Williamson, MOD5 User's Manual, NPS-61WN71061A,
1971.
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NESC0491/01
source program mag tape SRCTPtest-case data mag tape DATTP
test-case output mag tape OUTTP
report REPPT
Keywords: multigroup, scattering, slowing-down, spectra, stochastic processes, time dependence.