The reprocessing of used nuclear fuel requires the dissolution and separation of numerous radioisotopes that are present as fission products in the fuel. The leading technology option in the U.S. for reprocessing is a sequence of processing methods known as UREX+ (Uranium Extraction ). However, an industrial scale facility implementing this separation procedure will require the establishment of safeguards and security systems to ensure the protection of the separated materials. A number of technologies have been developed for meeting the measurement demands for such a facility. This project focuses on the design of a gamma detection system for taking measurements of the flow streams of such a reprocessing facility. An experimental apparatus was constructed capable of pumping water spiked with soluble radioisotopes under various flow conditions through a stainless steel coil around a sodium iodide (NaI) detector system. Experiments were conducted to characterize the impact of flow rate, pipe air voids, geometry, and radioactivity dilution level on activity measurements and gamma energy spectra. Two coil geometries were used for these experiments, using 0.5 in stainless steel pipe wound into a coil with a 6 inch diameter; the first coil was 5.5 revolutions tall and the second coil was 9.5 revolutions tall. The isotopes dissolved in the flowing water were produced at the Texas A&M Nuclear Science Center via neutron activation of chromium, gold, cerium, and ytterbium nitrate salts. After activation, the salts were dissolved in distilled water and inserted into the radioactive flow assembly for quantitative measurements. Flow rate variations from 100 to 2000 ml/min were used and activity dilution levels for the experiments conducted were between 0.02 and 1.6 μCi/liter. Detection of system transients was observed to improve with decreasing flow rate. The detection limits observed for this system were 0.02 μCi/liter over background, 0.5% total activity change in a pre-spiked system, and a dilution change of 2% of the coil volume. MCNP (Monte Carlo N-Particle Transport) models were constructed to simulate the results and were used to extend the results to other geometries and piping materials as well as simulate actual UREX stream material in the system. The stainless steel piping for the flow around the detector was found to attenuate key identifying gamma peaks on the low end of the energy spectrum. For the proposed schedule 40 stainless steel pipe for an actual reprocessing facility, gamma rays below 100 keV in energy would be reduced to less than half their initial intensities. The exact ideal detection set up is largely activity and flow stream dependant. However, the characteristics best suited for flow stream detection are: 1) minimize volume around detector, 2) low flow rate for long count times, and 3) low attenuation piping material such as glass.

Reprocessing nuclear fuel is becoming more viable in the United States due to the anticipated increase in construction of nuclear power plants, the growing stockpile of existing used nuclear fuel, and a public desire to reduce the amount of this fuel. However, a new reprocessing facility in non-weapon states must be safeguarded and new reprocessing facilities in weapon states will likely have safeguards due to political and material accountancy reasons. These facilities will have state of the art controls and monitoring methods to safeguard special nuclear materials, as well as to provide real-time monitoring. The focus of this project is to enable the development of a safeguards strategy that uses well established photon measurement methods to characterize samples from the UREX+3a reprocessing method using a variety of detector types and measurement times. It was determined that the errors from quantitative measurements were too large for traditional safeguards methods; however, a safeguards strategy based on qualitative gamma ray and neutron measurements is proposed. The gamma ray detection equipment used in the safeguard strategy could also be used to improve the real-time process monitoring in a yet-to-be built facility. A facility that had real-time gamma detection equipment could improve product quality control and provide additional benefits, such as waste volume reduction. In addition to the spectral analyses, it was determined by Monte Carlo N Particle (MCNP) simulations that there is no noticeable self shielding for internal pipe diameters less than 2 inches, indicating that no self shielding correction factors are needed. Further, it was determined that HPGe N-type detectors would be suitable for a neutron radiation environment. Finally, the gamma ray spectra for the measured samples were simulated using MCNP and then the model was extended to predict the responses from an actual reprocessing scenario from UREX+3a applied to fuel that had a decay time of three years. The 3-year decayed fuel was more representative of commercially reprocessed fuel than the acquired UREX+3a samples. This research found that the safeguards approach proposed in this paper would be best suited as an addition to existing safeguard strategies. Real-time gamma ray detection for process monitoring would be beneficial to a reprocessing facility and could be done with commercially available detectors.

As nuclear energy grows in the United States and around the world, the expansion of the nuclear fuel cycle is inevitable. All currently deployed commercial reprocessing plants are based on the Plutonium - Uranium Extraction (PUREX) process. However, this process is not implemented in the U.S. for a variety of reasons, one being that it is considered by some as a proliferation risk. The 2001 Nuclear Energy Policy report recommended that the U.S. "develop reprocessing and treatment technologies that are cleaner, more efficient, less waste-intensive, and more proliferation-resistant." The Uranium Extraction (UREX+) reprocessing technique has been developed to reach these goals. However, in order for UREX+ to be considered for commercial implementation, a safeguards approach is needed to show that a commercially sized UREX+ facility can be safeguarded to current international standards. A detailed safeguards approach for a UREX+1a reprocessing facility has been developed. The approach includes the use of nuclear material accountancy (MA), containment and surveillance (C/S) and solution monitoring (SM). Facility information was developed for a hypothesized UREX+1a plant with a throughput of 1000 Metric Tons Heavy Metal (MTHM) per year. Safeguard goals and safeguard measures to be implemented were established. Diversion and acquisition pathways were considered; however, the analysis focuses mainly on diversion paths. The detection systems used in the design have the ability to provide near real-time measurement of special fissionable material in feed, process and product streams. Advanced front-end techniques for the quantification of fissile material in spent nuclear fuel were also considered. The economic and operator costs of these systems were not considered. The analysis shows that the implementation of these techniques result in significant improvements in the ability of the safeguards system to achieve the objective of timely detection of the diversion of a significant quantity of nuclear material from the UREX+1a reprocessing facility and to provide deterrence against such diversion by early detection.