**1. NAME AND TITLE**

O6R: A General-Purpose Monte Carlo Transport Code System.

**AUXILIARY ROUTINES**

O6R-ACT1FK: Simultaneous Neutron Transport and Analysis.

NUSECT: Cross Section Handling Code Package - ENDF Format.

EDIT: Data Converter - ENDF Format - BCD to Binary Mode.

GEOM: Generalized Geometry.

PICTURE: Geometry Input Diagnostic Code.

**NON-STANDARD LIBRARY SUBROUTINES**

Cylindrical Geometry.

Slab Geometry.

Spherical Geometry.

The O6R code system represents early development of the Oak Ridge Monte Carlo Code series which began with O5R (CCC-17) and later resulted in the MORSE series. This package is archived, retained for reference purposes, and is no longer recommended for routine use.

**2. CONTRIBUTOR**

Oak Ridge National Laboratory, Oak Ridge, Tennessee.

**3. CODING LANGUAGE AND COMPUTER**

FORTRAN IV and Assembler Language; IBM 360/75/91.

**4. NATURE OF PROBLEM SOLVED**

The O6R code system consists of modifications and extensions made to CCC-17/O5R, designed to calculate any quantity related to neutron transport in reactor or shielding problems. The system solves the integral Boltzmann transport equation in general phase space for the collision density of neutrons in the system. By using the appropriate analysis routines, quantities of interest such as the flux, absorption rate, neutron lifetime, and slowing down density can be estimated. A wide variety of problems in the areas of radiation shielding, reactor analysis and design, criticality safety, and neutron scattering experiments can be solved with the O6R system.

The O6R-ACT1FK prototype permits simultaneous neutron transport and analysis, the specific features slanted toward the application to calculations of shields rather than reactors. It embodies all the features of O5R and ACT1FK-analysis codes with substantial improvements.

**5. METHOD OF SOLUTION**

O6R solves the integral form of the Boltzmann equation by the Monte Carlo method. It is based on the CCC-17/O5R General Purpose Monte Carlo code system and employs the same general method of solution. New features have been incorporated. Rather than having cross-section information for only one supergroup in core, O6R allows information for as many supergroups as can be accommodated by computer memory. In addition, in-core analysis of neutron histories can be performed. A parameter, AGE, is computed and stored for use in the calculation of time-dependent problems. An option allows a medium to be treated as an albedo medium. A user routine is called to pick a new energy and direction. Path stretching, a variance reduction technique equivalent to the exponential transform, has been made a standard feature. A user-written routine is incorporated to calculate inelastic scatter. ENDF/B formatted cross-section data can be used to generate cross section input to O6R using NUSECT. Gamma-ray generation during neutron transport is also available. Subroutine GAMMA generates a tape containing appropriate parameters. The user supplies subroutines SETGAM and GAMGEN.

**6. RESTRICTIONS AND LIMITATIONS**

The O6R system is presently limited to 16 media in the cross section representation. This can be increased by changing a dimension statement. The order of the scattering is limited to a P16 Legendre expansion.

**7. TYPICAL RUNNING TIME**

Running time for O6R on various types of problems varies extremely from a few seconds up to an hour on the IBM 360/91.

Estimated IBM 360 CPU time in seconds of the packaged sample problems: EDIT, 8.43; NUSECT, 44.50; O6R, 51.50; and PICTURE, 8.44.

**8. COMPUTER HARDWARE REQUIREMENTS**

Uses standard I-O and a maximum of 3 tapes or direct access devices. Maximum region size used is 474 K.

**9. COMPUTER SOFTWARE REQUIREMENTS**

O6R is operable on the IBM 360/91 operating system using the OS-360 FORTRAN H (Level 18) compiler.

**10. REFERENCES**

C. L. Thompson and E. A. Straker, "O6R-ACT1FK, Monte Carlo Neutron Transport Code," ORNL-TM-3050 (August 1969).

D. C. Irving, "Changes to ORNL-3622," Informal Memos (1967).

R. R Coveyou, J. G. Sullivan, H. P. Carter, D. C. Irving, R. M. Freestone, Jr., and F. B. K. Kam, "O5R, A General-Purpose Monte Carlo Neutron Transport Code," ORNL-3622 (February 1965).

D. C. Irving, V. R. Cain, and R. M. Freestone, Jr., "An Amplification of Selected Portions of the 05R Monte Carlo Code User's Manual," ORNL-TM-2601 (May 1969).

F. B. K. Kam and K. D. Franz, "ACT1FK, A General Analysis Code for O5R," ORNL-3856 (September 1966).

D. C. Irving, "Description of the CDC-1604 Version of the O6R Neutron Monte Carlo Transport Code," ORNL-TM-3458 (May 1971).

D. C. Irving and G. W. Morrison, "PICTURE: An Aid in Debugging GEOM Input Data," ORNL-TM-2892 (May 1970).

E. A. Straker and V. R. Cain, "A Note on Subroutine CEASE (05R:EVAP), A Modification of EVAP in O5R," ORNL-TM-1552 (August 1966).

C. E. Burgart, "Treatment of 6Li(n,dn)alpha, 7Li(n,tn)alpha Reactions in O6R Random Walks and Analysis," ORNL-TM-3051 (April 1970).

C. L. Thompson, "Description of the 06R-ACT1FK Sample Problem and of EDIT," Informal Notes (November 1969).

**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 and output.

**12. DATE OF ABSTRACT**

August 1971; revised December 1984 and September 1991.

**KEYWORDS: ** MONTE CARLO; COMPLEX GEOMETRY; NEUTRON; TIME-DEPENDENT; ENDF/B FORMAT