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.

In the event that a radiological dispersal device (RDD) is detonated in the U.S. or near U.S. interests overseas, it will be crucial that the actors involved in the event can be identified quickly. If irradiated nuclear fuel is used as the dispersion material for the RDD, it will be beneficial for law enforcement officials to quickly identify where the irradiated nuclear fuel originated. One signature which may lead to the identification of the spent fuel origin is the isotopic composition of the RDD debris. The objective of this research was to benchmark a forward model methodology for predicting isotopic composition of spent nuclear fuel used in an RDD while at the same time optimizing the fidelity of the model to reduce computational time. The code used in this study was Monteburns-2.0. Monteburns is a Monte Carlo based neutronic code utilizing both MCNP and ORIGEN. The size of the burnup step used in Monteburns was tested and found to converge at a value of 3,000 MWd/MTU per step. To ensure a conservative answer, 2,500 MWd/MTU per step was used for the benchmarking process. The model fidelity ranged from the following: 2-dimensional pin cell, multiple radial-region pin cell, modified pin cell, 2D assembly, and 3D assembly. The results showed that while the multi-region pin cell gave the highest level of accuracy, the difference in uncertainty between it and the 2D pin cell (0.07% for 235U) did not warrant the additional computational time required. The computational time for the multiple radial-region pin cell was 7 times that of the 2D pin cell. For this reason, the 2D pin cell was used to benchmark the isotopics with data from other reactors. The reactors from which the methodology was benchmarked were Calvert Cliffs Unit #1, Takahama Unit #3, and Trino Vercelles. Calvert Cliffs is a pressurized water reactor (PWR) using Combustion Engineering 14??14 assemblies. Takahama is a PWR using Mitsubishi Heavy Industries 17??17 assemblies. Trino Vercelles is a PWR using non-standard lattice assemblies. The measured isotopic concentrations from all three of the reactors showed good agreement with the calculated values.

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The safety case for a proposed geological disposal facility (GDF) for radioactive wastes relies upon a series of engineered and natural barrier systems to limit the migration of harmful radionuclides into the geosphere over geological timescales. Natural minerals, dominantly phyllosilicates, are expected to be the most reactive components of both the host rock and the clay-based backfill surrounding the highly radioactive waste canisters for as long as 100,000 years. Upon eventual canister degradation, alpha-emitting radionuclides will leach into the backfill material (and eventually beyond) and the constituent mineral systems will accumulate radiation damage upon radionuclide uptake and/or surface precipitation. The following study is an assessment of the structural and chemical effects caused by alpha-particle bombardment of silicate minerals, as proxies for the radiation stability of natural materials present in the near and far field of a GDF.Microscopy and spectroscopy studies from naturally occurring radiation damage accumulated in silicates over geological timescales (forming distinct 'radiohaloes') have shown that both alpha-particle and alpha-recoil bombardment results in altered unit cell dimensions caused by the accumulation of point (Frenkel) defects. In the example of highly damaged biotite, structural breakdown through the reorientation of discrete lattice crystallites was observed; the variability of the interlayer spacing within these regions reveal the potential for damaged mica to adopt the structure of phyllosilicate breakdown products over geological time. Controlled alpha-particle irradiation using the Dalton Cumbrian Facility's 5 MV tandem pelletron ion accelerator, combined with microfocus spectroscopy analysis has revealed the mechanisms of high fluence alpha-radiation damage across 2:1 phyllosilicate minerals (biotite and chlorite); reducing the layered structures into a series of loosely connected domains of alternating lattice expansion and collapse. Radiation induced Fe redox changes have been revealed, with Fe reduction apparent at relatively low alpha-particle doses, giving way to Fe oxidation at high doses. A 'redox gradient', based on alpha-particle energy deposition through a silicate structure has therefore been proposed. In addition, the increase in 'edge' sites generated by structural deformation has been shown to be favourable for the adsorption of the Se(IV) oxyanion to the mica surface. Comprising a body of additional work, a core sample has been extracted from a spent nuclear fuel pond wall at the decommissioned Hunterston A nuclear power station and the radioactive contamination on the painted core surface has been analysed by microfocus spectroscopy. The contaminant radiostrontium has been shown to be associated with the Ti rich pigment in the surface paint, resulting in a 'patchy' accumulation of radioactivity at the core surface. In addition, inert Cs reactivity experiments using the underlying concrete have shown that Cs is preferentially uptaken by phyllosilicates within the altered mafic clasts used in the concrete aggregate.

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The safeguards verification of spent nuclear fuel is one of the major concern for the safeguards community, as this material represents about 80% of all material placed under safeguards.This PhD thesis described the development of two passive non-destructive assay (NDA) techniques: the Self-Indication Neutron Resonance Densitometry (SINRD) and the Partial Defect Tester (PDET).The NDA methods were investigated with Monte Carlo simulations and the benchmark experiments for SINRD were performed at the GELINA facility of JRC-IRMM in Geel, Belgium.The results for the SINRD technique showed promising results for the direct quantification of 239Pu in spent fuel, and both techniques gave encouraging results for the detection of partial defects. / Doctorat en Sciences de l'ingénieur et technologie / info:eu-repo/semantics/nonPublished