NEA-0573 LASER-PNC.
LASER-PNC, Neutron Spectra in Uniform Lattice with Burnup Calculation
NAME OR DESIGNATION OF PROGRAM, COMPUTER, NATURE OF PHYSICAL PROBLEM SOLVED, METHOD OF SOLUTION, RESTRICTIONS ON THE COMPLEXITY OF THE PROBLEM, TYPICAL RUNNING TIME, FEATURES, RELATED AND AUXILIARY PROGRAMS, STATUS, REFERENCES, MACHINE REQUIREMENTS, LANGUAGE, OPERATING SYSTEM OR MONITOR UNDER WHICH PROGRAM IS EXECUTED, OTHER RESTRICTIONS, NAME AND ESTABLISHMENT OF AUTHOR, MATERIAL, CATEGORIES
[ top ]
[ top ]
To submit a request, click below on the link of the version you wish to order.
Only liaison officers are authorised to submit online requests. Rules for requesters are
available here.
| Program name | Package id | Status | Status date |
|---|---|---|---|
| LASER-PNC | NEA-0573/01 | Tested | 01-DEC-1979 |
Machines used:
| Package ID | Orig. computer | Test computer |
|---|---|---|
| NEA-0573/01 | IBM 3033 | IBM 3033 |
[ top ]
3. NATURE OF PHYSICAL PROBLEM SOLVED
The program computes the neutron spectrum in a uniform lattice made up of cylindrical rods, cladding, and surrounding moderator. The LASER is based on modified versions of the slowing-down program MUFT and the thermalization transport theory program THERMOS. The program performs a burnup calculation for the lattice. The spatial distribution of burnup within the fuel rods is explicitly calculated. This calculation accounts for the variation of the neutron flux in space and energy during each time step. A buckling and a boron poison search (criticality search) are provided as options.
The program computes the neutron spectrum in a uniform lattice made up of cylindrical rods, cladding, and surrounding moderator. The LASER is based on modified versions of the slowing-down program MUFT and the thermalization transport theory program THERMOS. The program performs a burnup calculation for the lattice. The spatial distribution of burnup within the fuel rods is explicitly calculated. This calculation accounts for the variation of the neutron flux in space and energy during each time step. A buckling and a boron poison search (criticality search) are provided as options.
[ top ]
4. METHOD OF SOLUTION
The methods used in solving the neutron transport equation are essentially those utilized by the MUFT and THERMOS programs. The depletion equations are solved by assuming a polynomial expansion for exponential functions. The non-linear effects in the burn-up equations are accounted for by computing the rate of change of the neutron flux with time and by using the Runge-Kutta numerical procedure to recover the flux as a function of time. The production and loss of chain members during irradiation are evaluated by simple matrix algebra.
The methods used in solving the neutron transport equation are essentially those utilized by the MUFT and THERMOS programs. The depletion equations are solved by assuming a polynomial expansion for exponential functions. The non-linear effects in the burn-up equations are accounted for by computing the rate of change of the neutron flux with time and by using the Runge-Kutta numerical procedure to recover the flux as a function of time. The production and loss of chain members during irradiation are evaluated by simple matrix algebra.
[ top ]
5. RESTRICTIONS ON THE COMPLEXITY OF THE PROBLEM
The maximum number of space points is 14, with a maximum of 5 points in the fuel region. The code is restricted to 4 mixtures (fuel, cladding, moderator, and non-absorbing heavy scatterer). The moderator can be either light water (Nelkin or free gas scattering kernel), or heavy water (Nelkin scattering kernel). The cladding material can be stainless steel, aluminium or zircaloy-2. The fuel can be a metal, oxide or cermet. The epithermal and fast energy ranges include 50 energy groups. The thermal range includes 35 energy groups.
The maximum number of space points is 14, with a maximum of 5 points in the fuel region. The code is restricted to 4 mixtures (fuel, cladding, moderator, and non-absorbing heavy scatterer). The moderator can be either light water (Nelkin or free gas scattering kernel), or heavy water (Nelkin scattering kernel). The cladding material can be stainless steel, aluminium or zircaloy-2. The fuel can be a metal, oxide or cermet. The epithermal and fast energy ranges include 50 energy groups. The thermal range includes 35 energy groups.
[ top ]
6. TYPICAL RUNNING TIME
Execution time for no-burnup cases are approximately 2 minutes when the standard THERMOS iteration technique is used without extrapolation. For burnup problems the execution time is approximately 4 minutes per time step (in case of CDC-6600, 1.5 minutes per time step) when the linear approximation to the burnup equation is used.
Execution time for no-burnup cases are approximately 2 minutes when the standard THERMOS iteration technique is used without extrapolation. For burnup problems the execution time is approximately 4 minutes per time step (in case of CDC-6600, 1.5 minutes per time step) when the linear approximation to the burnup equation is used.
[ top ]
[ top ]
[ top ]
10. REFERENCES
- T. Kajiyama, R. Yumoto:
Improvement of Burnup Analysis Code (LASER) In Lattice Cell of Light Water Moderated Reactor, PNCT831-74-02, Tokai Works Semi-Annual Progress Report (1974).
- T. Kajiyama:
User's Manual for LASER-PNC (Draft), (June 1979).
- T. Kajiyama, R. Yumoto:
Improvement of Burnup Analysis Code (LASER) In Lattice Cell of Light Water Moderated Reactor, PNCT831-74-02, Tokai Works Semi-Annual Progress Report (1974).
- T. Kajiyama:
User's Manual for LASER-PNC (Draft), (June 1979).
NEA-0573/01, included references:
- G.C. Poncelet:Burnup Physics of Heterogeneous Reactor Lattices
WCAP-6069, UC-34 (June 1965).
- G.C. Poncelet:
LASER - A Depletion Program for Lattice Calculations Based on MUFT
and THERMOS
WCAP-6073, UC-32 (April 1966).
- LASER, ACC No. 249.360
ACC Programming Note 74-7 (October 29, 1973).
[ top ]
[ top ]
[ top ]
[ top ]
[ top ]
NEA-0573/01
| File name | File description | Records |
|---|---|---|
| NEA0573_01.001 | JCL AND INFORMATIONS | 65 |
| NEA0573_01.002 | ENDF/B LIBRARY DATA | 14770 |
| NEA0573_01.003 | LIBRARY PREPARATION PROGRAM | 142 |
| NEA0573_01.004 | LASER-PNC SOURCE (F4,EBCDIC) | 7251 |
| NEA0573_01.005 | SAMPLE INPUT FOR LASER-PNC | 19 |
| NEA0573_01.006 | PRINTED OUTPUT OF LIB. PREP. PROGRAM | 62 |
| NEA0573_01.007 | PRINTED OUTPUT OF LASER-PNC | 17712 |
| NEA0573_01.008 | PUNCHED OUTPUT OF LASER-PNC | 486 |
[ top ]
- B. Spectrum Calculations, Generation of Group Constants and Cell Problems
- D. Depletion, Fuel Management, Cost Analysis, and Power Plant Economics
Keywords: buckling, burnup, cladding, criticality searches, depletion, fuel rods, moderators, neutron spectra, reactor lattices, slowing-down, transport theory.