1. NAME AND TITLE
LASER: A One-Dimensional, Neutron-Thermalization, Lattice-Cell Program Based on MUFT
LASER THERMAL LIBRARY.
Westinghouse Electric Corporation, Pittsburgh, Pennsylvania.
Electric Power Research Institute, Palo Alto, California.
3. CODING LANGUAGE AND COMPUTER
Fortran IV; IBM 360/370.
4. NATURE OF PROBLEM SOLVED
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 search (criticality search) are provided as options. Output includes edits
in the energy range less than or equal to 0.625 eV.
5. METHOD OF SOLUTION
The methods used in solving the neutron transport equation are essentially those utilized by the
MUFT and THERMOS programs. The non-linear 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.
6. RESTRICTIONS OR LIMITATIONS
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 (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, 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 input.
7. 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 1-2 minutes when the maximum number of
space points is specified. For a burnup problem, execution time is 1-2 minutes when the linear
approximation is chosen, and 4-6 minutes when non-linear effects are included.
8. COMPUTER HARDWARE REQUIREMENTS
The code is operable on the IBM 360/370 with library tape and 2 scratch units. It requires 150
9. COMPUTER SOFTWARE REQUIREMENTS
A Fortran IV compiler and a standard operating system are required.
Argonne Code Center Programming Note 74-4.
B. R. Leonard, Jr., D. A. Kottwitz, U. P. Jenquin, K. B. Stewart, and C. M. Heeb, "Cross-Section Standardization for Thermal Power Reactors," EPRI 221 (July 1975).
C. G. Poncelet, "LASER - A Depletion Program for Lattice Calculations Based on MUFT and THERMOS," WCAP-6073 (April 1966).
C. G. Poncelet, "Burnup Physics of Heterogeneous Reactor Lattices," WCAP-6069 (June 1965).
11. CONTENTS OF CODE PACKAGE
Included are the referenced documents and one (1.2MB) DOS diskette which contains the source
codes and sample problem input.
12. DATE OF ABSTRACT
KEYWORDS: ONE-DIMENSION; CYLINDRICAL GEOMETRY; BURNUP; NEUTRON; FISSION PRODUCTS; THERMALIZATION; CELL CALCULATION