BURNUP: A FORTRAN IV CODE FOR COMPUTING U AND Pu FUEL BURNUP FROM U, Pu, AND Nd MASS SPECTROMETRIC MEASUREMENTS.

Abstract

The computer code BURNUP was written in 1964 and has been in use at this laboratory since that time. During this three-year period of use, it has been debugged and improved. The present form represents the sixth production version. Although a single sample can be processed, the averaging procedures for nuclear constants are most effective when (ive to fifty samples from the same fuel load are pooled in a single case. The code will process data obtained by the American Society for Testing and Materials

Isodil: A Computer Code for Processing Isotopic Dilution Measurements From Spent Fuel Samples.

“…Optional output features include plots of measured data from a series of samples (these plots are often useful for detecting anomalous data) and punched input cards for the DBUFIT(l) least squares analysis code. Other codes with some similarities to ISODIL are REBUF (2) and BURNUP (3).…”

Evaluation of the irradiation-averaged fission yield for burnup determination in spent fuel assays

“…Coupled neutron transport and depletion calculations are generally validated against well documented spent nuclear fuel (SNF) assay data. Those assays often report, in addition to the spent nuclear fuel inventory, the experimental sample burnup [1][2][3][4][5][6][7][8][9][10][11][12]. The experimental burnup offers a simple way to cross-check the burnup or the power history calculated by the plant operator.…”