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A Generalized Method For Fissile Material Characterization Using Short-Lived Fission Product Gamma Spectroscopy, Justin Richard Knowles

Doctoral Dissertations

Characterizing the fissile content of nuclear materials is of particular interest to the safeguards and nuclear forensics communities. Short-lived fission product gamma spectroscopy offers a significant reduction in analysis time and detection limits when compared to traditional non-destructive assay measurements. Through this work, a fully generalizable method that can be applied to variations in fissile compositions and neutron spectra was developed for the modeling and measurement of short-lived fission product gamma-rays. This method uses a 238-group neutron flux that was characterized for two pneumatic tube positions in the High Flux Isotope Reactor using flux monitor irradiations. This flux spectrum was …

Process Attribute Monitoring Approach To Gaseous Centrifuge Enrichment Plant Safeguards, Adam Michael Shephard

Doctoral Dissertations

Unattended and remote monitoring systems for the early detection of gas centrifuge enrichment plant (GCEP) misuse have long been sought by the International Atomic Energy Agency (IAEA) safeguards inspectorate, but have yet to be realized for most GCEPs. The primary misuse of concern is high enriched uranium (HEU) production but additionally include excess low enriched uranium (LEU) production and material diversion.

The current research is a feasibility study that investigates a process monitoring approach combined with passive gamma spectroscopy techniques to monitor GCEPs. In the current research, a simulated reference cascade and monitoring system is used to demonstrate several techniques …

Advancement Of Computational Nuclear Fuel And Safeguard Tools, Nathan Shoman

Masters Theses

This thesis work was conducted in two separate but related areas: reactor physics and nuclear fuel safeguards. Reactor physics calculations at the assembly level are important to understand the buildup of fission products and major actinides. These calculations can be used to determine the content of a spent fuel assembly prior to reprocessing or disposal at a much cheaper price. This thesis presents an extension to the SCALE (A Nuclear Systems Modeling & Simulations Suite) by developing a lattice model and cross-section library for the CANada Deuterium Uranium (CANDU) 37 and 28 element assemblies. The previous cross-section library was based …