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NESC0249 LASER

LASER, Slowing-Down Neutron Spectra and Burnup for Thermal Reactors, Neutron Transport Theory

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1. NAME OR DESIGNATION OF PROGRAM:  LASER
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2. COMPUTERS
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Program name Package id Status Status date
LASER NESC0249/01 Tested 01-APR-1977

Machines used:

Package ID Orig. computer Test computer
NESC0249/01 IBM 370 series IBM 370 series
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3. DESCRIPTION OF PROBLEM OR FUNCTION

LASER is based on modified versions of the slowing-down program MUFT and the thermalization transport theory program THERMOS, and performs a calculation of the  neutron spectrum in a uniform lattice made up of cylindrical rods, cladding, and surrounding moderator. The thermal cutoff in LASER is  1.855 eV. The program performs a burnup calculation for the lattice. The spatial distribution of burnup within the fuel rods is explicitly calculated. The program will, at option, account for all  non-linearities and mutual connections in the system of burnup equations. This calculation accounts for the variation of the neutron flux in space and energy during each time-step. A buckling,  and a boron poison, criticality search are provided as options. Output includes edits in the energy range 0 to 0.625 eV.
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4. METHOD OF SOLUTION

The methods used in solving the neutron transport equation are essentially those utilized by the MUFT and THERMOS programs. The nonlinear effects in the burnup 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 depletion equations are  solved by assuming a polynomial expansion for exponential functions. The production and loss of chain members during irradiation are evaluated by simple matrix algebra. The procedure allows for a time- dependent flux in the form of a power series.
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5. RESTRICTIONS ON THE COMPLEXITY OF THE PROBLEM

This version of LASER is restricted to one-dimensional, cylindrical geometry. The maximum number of space points is 14, with a maximum of 5 space points in the fuel region. The code is restricted to 4 mixtures (moderator, non-absorbing heavy scatter, cladding, fuel). 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, aluminum or zircalloy-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. Only the U235 chain (through U236) and the U238 chain (through Pu242) are available in the code. The fission products are  separated into Xe135, the directly-produced Sm149, and all other fission products lumped into one pseudo fission product. The cross sections for the lumped fission products are represented by polynomials in the burnup. The spatial distribution of U238 resonance captures within the fuel rods is given as input.
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6. TYPICAL RUNNING TIME

The greater part of the LASER execution time  is consumed by the THERMOS calculation. Execution times for no- burnup cases are approximately 1 to 2 minutes when the maximum number of space points is specified. For a burnup problem, execution time is 1 to 2 minutes when the linear approximation is chosen, and  4 to 6 minutes when non-linear effects are included.
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7. UNUSUAL FEATURES OF THE PROGRAM

The program provides a complete edit for the lattice in the energy range 0 to 10 MeV, including integral quantities and homogenized few-group cross sections. The user has a choice of thermal neutron scattering kernels, resonance self-shielding treatments, and cell criticality searches. Linear and nonlinear treatments of the burnup equations are provided. The detailed spatial burnup distribution within the fuel rods is taken into account. A burnup problem can be continued after any time-step, by using the special input deck punched after each time-step.
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8. RELATED AND AUXILIARY PROGRAMS

The original IBM7094 version of LASER was prepared by C.G. Poncelet of the Westinghouse PWR Plants Division, Pittsburgh, Pennsylvania. LIBP generates the thermal library tape. SIG1 converts, by interpolation, the cross section data corresponding to a given energy mesh to cross section data suitable for input to the LIBP program.
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9. STATUS
Package ID Status date Status
NESC0249/01 01-APR-1977 Tested at NEADB
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10. REFERENCES:
NESC0249/01, included references:
- C.G. Poncelet:
  Burnup Physics of Heterogeneous Reactor Lattices
  WCAP-6069, UC-34 (June 1965).
- C.G. Poncelet:
  LASER - A Depletion Program for Lattice Calculations Based on MUFT
  and THERMOS
  WCAP-6073, UC-32 (April 1966).
- ACC Programming  Note 74-7:
  LASER, ACC No. 249.360 (October 29, 1973).
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11. MACHINE REQUIREMENTS:  150K with library tape and 2 scratch units.
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12. PROGRAMMING LANGUAGE(S) USED
Package ID Computer language
NESC0249/01 FORTRAN-IV
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13. OPERATING SYSTEM UNDER WHICH PROGRAM IS EXECUTED:   OS/360.
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14. OTHER PROGRAMMING OR OPERATING INFORMATION OR RESTRICTIONS:
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15. NAME AND ESTABLISHMENT OF AUTHORS

                 C. E. Carson
                 Nuclear Engineering Branch
                 Tennessee Valley Authority
                 303 Power Building
                 Chattanooga, Tennessee  37401
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16. MATERIAL AVAILABLE
NESC0249/01
File name File description Records
NESC0249_01.001 INFORMATION 2
NESC0249_01.002 LIBRARY PREPARATION PROGRAM - F4 EBCDIC 385
NESC0249_01.003 LASER SOURCE - FORTRAN IV EBCDIC 7783
NESC0249_01.004 BCD LIBRARY 1496
NESC0249_01.005 SAMPLE PROBLEM INPUT 15
NESC0249_01.006 SAMPLE PROBLEM OUTPUT 1154
NESC0249_01.007 JOB CONTROL CARDS 18
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17. CATEGORIES
  • B. Spectrum Calculations, Generation of Group Constants and Cell Problems
  • D. Depletion, Fuel Management, Cost Analysis, and Power Plant Economics

Keywords: criticality searches, cylinders, depletion, one-dimensional, reactor lattices, slowing-down, spectra, thermalization.