Verifying the production of fissile material in nuclear facilities is a key element in the efforts of nuclear nonproliferation. The production of fissile material can result from several processes. Some of these processes include the enrichment of uranium for fuel to power nuclear reactors, the production of plutonium within the fuel of a reactor during operation, and the buildup of 233U from 232Th via the irradiation of thorium. During these processes some evidence is left behind that could lead an investigator to predict the most likely events from the past that would have led to the evidence observed in the present day. This study of evidence and its relationship to past material production is often known as “nuclear archaeology” and is rooted in the verification of nuclear weapons activities. Several methods have been developed previously for estimating cumulative energy production and plutonium production from graphite-moderated reactors. The Graphite Isotope Ratio Method (GIRM) is one well-known technique. This method is based on the measurement of trace isotopes in the reactor’s graphite matrix to determine the change in their isotopic ratios due to burnup. These measurements are then coupled with reactor calculations to determine the total plutonium and energy production of the reactor. To facilitate sensitivity analysis of these methods, a one-group cross section and fission product yield library for the fuel and graphite activation products has been developed for MAGNOX-style reactors. This library is intended for use in the ORIGEN computer code, which calculates the buildup, decay, and processing of radioactive materials. The library was developed using a fuel cell model in Monteburns. This model consisted of a single fuel rod including natural uranium metal fuel, magnesium cladding, carbon dioxide coolant, and Grade A United Kingdom (UK) graphite. Using this library a complete sensitivity analysis can be performed for GIRM and other techniques. The sensitivity analysis conducted in this study assessed various input parameters including 235U and 238U cross section values, aluminum alloy concentration in the fuel, and initial concentrations of trace elements in the graphite moderator. The results of the analysis yield insight into the GIRM method and the isotopic ratios the method uses as well as the level of uncertainty that may be found in the system results.
- K. E. Chesson, "A One-Group Parametric Sensitivity Analysis for the Graphite Isotope Ratio Method and Other Related Techniques Using ORIGEN 2.2", M.S. Thesis, Texas A & M University, August 2007.
- K.E. Chesson W.S. Charlton, "A Brief Sensitivity Analysis for the GIRM and Other Related Techniques using a One-Group Cross Section Library for Graphite-Moderated Reactors", Annual Meeting of the Institute for Nuclear Materials Management, July 8-12, 2007, Tucson, AZ.
- K.E. Chesson W.S. Charlton, "A Brief Sensitivity Analysis for the GIRM and Other Related Technique using a One-Group Cross Section Library for Graphite-Moderated Reactors", Proceedings of the 29th ESARDA Annual Meeting, Aix-en-Provence, France, May 22-24, 2007.