1. NAME AND TITLE

ESP: General Purpose Monte Carlo Neutron Transport Code System.

ESP is one of several codes developed at Oak Ridge National Laboratory which has the 05R code as its predecessor.

2. CONTRIBUTOR

Oak Ridge National Laboratory, Oak Ridge, Tennessee.

3. CODING LANGUAGE AND COMPUTER

FORTRAN IV and Assembler language; IBM 360/370.

4. NATURE OF PROBLEM SOLVED

ESP is a general purpose Monte Carlo reactor analysis code system. It covers the energy range 10+3 eV to 15 MeV and is designed for both eigenvalue and fixed source reactor and reactor cell calculations. Cross section and related nuclear data are accepted only in the version I, II, and III ENDF/B format. In addition to smooth data, the cross-section preparation includes a detailed treatment of resolved and unresolved resonances and the free gas and S(alpha,beta) thermal models. The collision and source routines utilize the ENDF/B nonelastic secondary distributions, the anisotropic scattering data, and the fission spectra. The energy range may be divided into as many as 25,000 intervals, and over each interval the appropriately averaged cross sections are used as constant point-energy values. A general three-dimensional geometry description is available, as well as several specialized geometries. Use of fixed-source options allows calculations on nonmultiplying systems. A steady-state analysis of neutron histories is performed in core providing such quantities as neutron fluxes, reaction rates, and cross sections, all averaged over arbitrary energy ranges and spatial regions.

5. METHOD OF SOLUTION

ESP utilizes standard Monte Carlo techniques in the neutron tracking and collision processes. Analysis of histories is performed by either collision density or track length estimation. The nuclides are assumed to be at rest except in the thermal energy range. Non-absorption weighting, splitting, and Russian roulette are used as variance reduction methods. To obtain estimates of the statistical error of ESP calculations, the neutron histories are processed in batches. The fission neutrons produced in a batch may be used as the starting neutrons for the next batch. Anisotropic scattering is handled by the Coveyou technique. The resonance cross sections are compiled using the Doppler broadening single-level Breit-Wigner model. A detailed statistical treatment is applied to this model in the unresolved resonance energy region.

6. RESTRICTIONS OR LIMITATIONS

The energy range is restricted to neutron energies for which data are available in the ENDF/B files. Each physical boundary of the system must be describable by a general second-order algebraic equation. There may be as many as 25 nuclides in as many as 16 different mixtures in the system.

7. TYPICAL RUNNING TIME

Running times are highly problem dependent. Less than 5 minutes were used on the IBM 360/91 for data preparation, tracking, and analysis of 20,000 histories for GODIVA. A heterogeneous mockup of ZPR-III Assembly 48 took 3 hours; 2 hours for a detailed resonance cross section data preparation, and 1 hour for the Monte Carlo calculation of 15,000 histories giving a 0.5 % standard deviation of the calculated eigenvalue. The IBM 360/75 time would be about 3.3 times slower.

8. COMPUTER HARDWARE REQUIREMENTS

ESP was designed for the IBM 360 computer using 14 direct access storage devices in addition to I-O. The sample problem (spherical geometry) used 480 K core storage.

9. COMPUTER SOFTWARE REQUIREMENTS

ESP has over 100 subroutines and approximately 16,000 source cards. A six-segment overlay structure is used to conserve computer storage. A library of cross sections in the ENDF/B format is required. Cross sections for MAT(102) and MAT(103), a Random Number Generator, and some non-standard library routines are included in the package.

10. REFERENCES

a. Included in the documentation:

S. N. Cramer, R. S. Carlsmith, G. W. Morrison, G. W. Perry, J. L. Lucius, "ESP: A General Purpose Monte Carlo Reactor Analysis Code," ORNL-TM-3164 (January 1972).

b. Background information:

D. C. Irving, R. M. Freestone, Jr., and F. B. K. Kam, "05R, A General Purpose Monte Carlo Neutron Transport Code," ORNL-3622 (February 1975).

S. N. Cramer, "Monte Carlo Analysis of the Exact Geometric Mockup of ZPR-II Assembly 48," ORNL-TM-3596 (October 1971).

11. CONTENTS OF CODE PACKAGE

Included are the referenced document (10.a) and one (1.2MB) DOS diskette which contains source programs, random number package, library routines, cross sections, and sample problem data and output.

12. DATE OF ABSTRACT

August 1975.

KEYWORDS: MONTE CARLO; NEUTRON; COMPLEX GEOMETRY; ENDF/B FORMAT; THERMALIZATION