The Graphite Isotope Ratio Method (GIRM) is a technique used to estimate the total plutonium production in a graphite-moderated reactor. The cumulative plutonium production in that reactor can be accurately determined by measuring neutron irradiation induced isotopic ratio changes in certain impurity elements within the graphite moderator. The method does not require detailed knowledge of a reactor's operating history, although that knowledge can decrease the uncertainty of the production estimate. The basic premise of the GIRM is that the fluence in non-fuel core components is directly related to the cumulative plutonium production in the nuclear fuel (Reid et al. 1999). This report provides a basic description of the physical basis and practical implementation of the Graphite Isotope Ratio Method. iii Glossary Burnup-The burnup is the total energy released by a given quantity of nuclear fuel in units of megawatt-days/metric ton (MWd/MT). Cross-Section-The total neutron cross-section is defined as the probability per unit path length traveled that a neutron will interact with a nucleus in the background media. There are specific cross sections for different reactions including scattering, capture, and fission. Cross-sections are typically denoted with the symbol Σ, and have units of inverse length (cm-1). In general, cross sections are complex functions of the neutron energy and the composition of the background media. See reaction rate and mean free path. Fissile-A fissile nuclide, such as U 235 or Pu 239 , can undergo fission following the absorption of a zeroenergy neutron. Fluence-The fluence is the time integral of the scalar flux in units of neutrons/cm 2. If the flux is constant in time, the fluence is simply the product of the flux and time. Flux-The flux is the product of the neutron density and the neutron speed with units of neutrons/cm 2second. The flux is often a very complex function of position (and energy). Mean free path (MFP)-The mean distance a neutron can be expected to travel between collisions with the background media. The mfp is equal to the inverse of the total cross-section (1/Σ). Moderator-Neutrons are produced in fission with very high energies. Fission cross-sections, however, are typically much larger at low neutron energies. Therefore, many nuclear reactors contain low atomic mass materials that serve to decrease average neutron energy through collisions. This material is referred to as moderator. Graphite, light water, and heavy water are frequently used as moderators. Reaction Rate-A reaction rate is the rate at which neutrons interact with the nuclei in the background media in units of reactions/cm 3-sec. It is the product the scalar flux and the cross-section for the reaction of interest. v
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