1. NAME AND TITLE

BCG: A Code For Calculating Pointwise Neutron Spectra and Criticality in Fast Reactor Cells.

AUXILIARY ROUTINE

SAR: An Interface Routine Between BCG and the Reconstructed Library from the LINEAR-RECENT-SIGMA1 Programs. DATA LIBRARY

TAPE9: Linearly Interpolable, Doppler Broadened Cross Section Library in ENDF/B-IV Format.

2. CONTRIBUTOR

Instituto De Estudos Avancados, Sao Jose Dos Campos, Sao Paulo, Brasil.

3. CODING LANGUAGE AND COMPUTER

FORTRAN IV; CDC 170/750.

4. NATURE OF PROBLEM SOLVED

BCG calculates the space-energy neutron flux distribution and effective multiplication factor of fast-reactor multiregion cylindrical cells. It makes no simplifying assumptions on mathematical models used in ENDF/B-IV to describe anisotropy and secondary neutron emission, thereby providing rigorous treatment of evaluated data.

5. METHOD OF SOLUTION

BCG solves iteratively a set of balance equations and interface current relations at each energy point in a grid which is the union of the individual cross section energy grids of the materials in the system and at an arbitrary number of annular zones.

6. RESTRICTIONS OR LIMITATIONS

Unbroadened average cross sections in the unresolved resonance region are given as input to BCG. In the present version, SAR has no provision for including temperature effects to the cross sections in this region. BCG accepts only materials with angular and energy distributions given by the laws used in the ENDF/B-IV library.

7. TYPICAL RUNNING TIME

On the CDC Cyber 170/750, the sample problem took 44 minutes. The test problem consists of a two-region (one zone and one material per region) critical fuel cell with an energy grid of 2,216 points. The problem required 28 iterations to converge within 1.0e-05 relative error in k-effective and 1.0e-04 maximum relative error in the total reaction rates.

8. COMPUTER HARDWARE REQUIREMENTS

BCG runs on the CDC Cyber 170/750.

9. COMPUTER SOFTWARE REQUIREMENTS

The code was written in FORTRAN IV, under the NOS 2.5 operating system.

10. REFERENCE

a. Included in the documentation:

S. Bogado Leite, A. D. Caldeira and R. D. M. Garcia. BCG: A Code for Calculating Pointwise Neutron Spectra and Criticality in Fast Reactor Cells, Report IEAv-001/88, Centro Tecnico Aeroespacial, Instituto de Estudos Avancados, Sao Paulo, Brasil, Feb. 1988.

A. D. Caldeira and R. D. M. Garcia, Update of the BCG Code Library, IEAv-012/90.

11. CONTENTS OF CODE PACKAGE

Included are the referenced document and three DS/HD (1.2 MB) 5.25-inch diskettes.

12. DATE OF ABSTRACT

January 1991, May 1991.