1. NAME AND TITLE
FPZD: Code System for Multigroup Neutron Diffusion/Depletion Calculations.
Marshall Space Flight Center, NASA, Huntsville, Alabama.
3. CODING LANGUAGE AND COMPUTER
FORTRAN 77; PC 386 and PC 486.
4. NATURE OF PROBLEM SOLVED
FPZD calculates the spatial power and neutron flux distributions of a nuclear reactor along with its effective multiplication factor and time-dependent material compositions.
5. METHOD OF SOLUTION
The depletion subroutines permit a very generalized nuclide depletion scheme to be used in which the materials may be distributed into a number of burnup subregions each containing a different atom density set. Depletion of the nuclides is calculated separately for each burnup subregion assuming exposure to the average flux of the spatial region. The macroscopic properties of a spatial region are calculated by giving the appropriate volume fraction weighting factors to the cross sections of the burnup subregions.
6. RESTRICTIONS OR LIMITATIONS
7. TYPICAL RUNNING TIME
On a 20 megahertz PC 386 with a math co-processor, the following times were noted: FPZDIN.INI took 45 minutes; FPZDIN.KIC took 25 minutes; FPZDIN.IC took 35 minutes; FPZDIN.KRL took 45 minutes; FPZDIN.RL took 40 minutes.
8. COMPUTER HARDWARE REQUIREMENTS
FPZD runs on the PC 386 or PC 486 under the DOS 5.0 operating system; a math co-processor is required.
9. COMPUTER SOFTWARE REQUIREMENTS
The code was written in FORTRAN 77 and was tested using the Microsoft Version 5.01 compiler.
B. Emmrich, "A One-Dimensional Multigroup Neutron Diffusion/Depletion Code for Performing Control and Fuel Loading Searches and for Determining the Fuel and Poison Distributions Required to Achieve Burnup Independent Power Distributions with Optional Thermal-Fluid Feedback," NASA informal document, Marshall Flight Center, Huntsville, Alabama.
11. CONTENTS OF CODE PACKAGE
Included are the referenced documents and 3 5.25-inch DS/HD (1.2MB) diskettes which include source and executable codes, sample input and output.
12. DATE OF ABSTRACT
KEYWORDS: BURNUP; DIFFUSION THEORY; NEUTRON; REACTOR PHYSICS; MICROCOMPUTER; MULTIGROUP; ONE-DIMENSION