Slides - Institiute of Nuclear Materials Management Annual Meeting (INMM), Atlanta GA, July 20-24, 2014

Slides - Institute of Nuclear Materials Management Annual Meeting, Atlanta GA, July 20-24, 2014

Slides - Institute of Nuclear Materials Management Annual Meeting (INMM), July 20-24 2014, Atlanta GA

A methodology for performing and applying nuclear criticality safety calculations, for PWR spent nuclear fuel (SNF) packages with actinide-only burnup credit, is described. The changes in the U-234, U-235, U-236, U-238, Pu-238, Pu-239, Pu-240, Pu-241, Pu-242, and Am-241 concentration with burnup are used in burnup credit criticality analyses. No credit for fission product neutron absorbers is taken. The methodology consists of five major steps. (1) Validate a computer code system to calculate isotopic concentrations of SNF created during burnup in the reactor core and subsequent decay.

In the 1980s, a series of critical experiments referred to as the Haut Taux de Combustion (HTC)
experiments was conducted by the Institut de Radioprotection et de Sûreté Nucléaire (IRSN) at the
experimental criticality facility in Valduc, France. The plutonium-to- uranium ratio and the isotopic
compositions of both the uranium and plutonium used in the simulated fuel rods were designed to be
similar to what would be found in a typical pressurized-water reactor fuel assembly that initially had an

Slides - Institute of Nuclear Materials Management, 55th Annual Meeting, July 20 – 24, 2014 Atlanta, Georgia

Slides - Institute of Nuclear Materials Management, 55th Annual Meeting, July 20 – 24, 2014 Atlanta, Georgia