NESC0572 XLACS.

1. NAME OR DESIGNATION OF PROGRAM:  XLACS.

3. NATURE OF PHYSICAL PROBLEM SOLVED

XLACS calculates fine-group averaged neutron cross sections from ENDF/B data. Its primary purpose is to produce full range multigroup libraries for the XSDRN  program (ACC Abstract 493). Provisions are included for treating fast, resonance, and thermal ENDF/B data. Fine-group energy structures and expansion orders used to represent differential cross sections for XSDRN can be arbitrarily set by the user. Cross sections can be averaged over an arbitrary input weighting function  or by any of several built-in functions.

4. METHOD OF SOLUTION

In the resolved resonance range XLACS accepts either single or multi-level Breit-Wigner parameters and calculates  infinite dilution and background cross sections consistent with the  Nordheim integral treatment used in XSDRN. Doppler broadening is provided through the use of PSI and CHI routines.
The unresolved treatment stems from MC**2 (ACC Abstract 355) but has been modified to treat the overlap of unresolved resonance sequences. Unresolved cross sections include some shielding effects  through a (1/sigma t) weighting and can include Doppler broadening.  Smooth cross sections are calculated by averaging the ENDF/B point data over either a user-supplied input spectrum or built-in spectra. Elastic scattering matrices are computed from the scattering angular distribution data specified as Legendre expansion coefficients. An inelastic scattering treatment is provided for all present forms of  ENDF/B file 5 data using either a discrete level or an evaporation model. Thermal cross section matrices are calculated through the use of procedures taken from the FLANGE2 code (ACC Abstract 368).

5. RESTRICTIONS ON THE COMPLEXITY OF THE PROBLEM

The principal restriction is the availability of adequate core storage to assimilate required arrays. The code is flexibly-dimensioned which means that array sizes are set for the particular problem at execution time. Core storage requirements are not affected by the number of nuclides in a case.

6. TYPICAL RUNNING TIME

Running time is a function of number of resonances, number of groups, Doppler broadening, and number of thermal scattering kernels being calculated. A U-238 calculation with 68 groups, no thermal calculation, and no Doppler broadening requires 1.65 minutes. A Pu-239 calculation with 123 groups (30 thermal), one thermal kernel, and Doppler broadening takes 6.72 minutes. The calculation of both Fe and H2O with 123 groups (30 thermal), and one thermal kernel requires 4.74 minutes.

7. UNUSUAL FEATURES OF THE PROGRAM

In contrast with most ENDF/B processing programs, the variable dimensioning technique employed by XLACS permits optimal use of available core storage. The program treats both fast and thermal ENDF/B data in a single calculation. Mixed ENDF/B tape modes (BCD and binary) are permitted. For convenience, editing, updating, and plotting capabilities are provided.

8. RELATED AND AUXILIARY PROGRAMS

When angular distributions of secondary neutrons are specified as tabular functions the SAD program must be used prior to running XLACS to convert the data to Legendre expansion coefficients. The XSDRN program uses the cross sections generated by XLACS. However, the XSDRN tape format produced by XLACS is somewhat more general than the tape format used by present XSDRN codes. An auxiliary program, convert, is available to  convert to the usable format.

10. REFERENCES

- N.M. Greene, J.L. Lucius, J.E. White, R.Q. Wright, C.W. Craven, Jr., and M.L. Tobias:
  XLACS, A Program to Produce Fine Group Averaged Neutron Cross Sections from ENDF/B.
  ORNL-TM-3646, April 3, 1972
- M.K. Drake:
  Editor, Data Formats and Procedures for The ENDF Neutron Cross Section Library.
  BNL-50279, October 1970
- N.M. Greene and C.W. Craven, Jr.:
  XSDRN, A Discrete Ordinates Spectral Averaging Code.
  ORNL-TM-2500, July 1969
- L.W. Nordheim:
  The Theory of Resonance Absorption.
  Symposium on Applied Mathematics, Vol. 11, 1961
- D.M. O'Shea, B.J. Toppel, and A.L. Rago:
  MC**2 - A Code to Calculate Multigroup Cross Sections.
  ANL-7318, June 1967
- H.C. Honeck and D.R. Finch:
  FLANGE-II, A Code to Process Thermal Neutron Data from an ENDF/B Tape
  DP-1278, October 1971

    - N. M. Greene, J. L. Lucius, J. E. White, Q. R. Wright,
      C. W. Craven, and M. L. Tobias;
      XLACS: A Program to Produce Weightend Multigroup Neutron Cross
      Sections from ENDF/B. ONRL-TM-3646 (AMPX-2) (April 3, 1972).

11. MACHINE REQUIREMENTS

330k bytes of core storage and external storage devices such as magnetic tapes or disks.

13. OPERATING SYSTEM OR MONITOR UNDER WHICH PROGRAM IS EXECUTED:  OS/360.

14. ANY OTHER PROGRAMMING OR OPERATING INFORMATION OR RESTRICTIONS

The program is presently used in a 3-level overlay structure consisting of 19 separate segments. Two functions, MODEL and ICLOCK, are not included with the program and should be dummied out.

15. NAME AND ESTABLISHMENT OF AUTHOR

                 N. M. Greene, J. L. Lucius, J. E. White,
                 R. Q. Wright, C. W. Craven, Jr., and M. L. Tobias
                 Oak Ridge National Laboratory
                 P. O. Box X
                 Oak Ridge, Tennessee 37830 U. S. A.

File name	File description	Records
NESC0572_02.001	SOURCE PROGRAM (FORTRAN)	12985
NESC0572_02.002	SAMPLE PROBLEMS (2 CASES)	47
NESC0572_02.003	JCL+OVERLAY CARDS	50
NESC0572_02.004	AUX. PROGRAM CONVERT SOURCE PROG.+JCL	593
NESC0572_02.005	OUTPUT LIST OF CASE 1	1953
NESC0572_02.006	OUTPUT LIST OF CASE 2	1766