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At a nuclear fuel reprocessing plant the special nuclear materials (SNM) are received as constituents of spent fuel assemblies, are converted to liquid form, and undergo a series of chemical processes. Uncertainties in measurements of SNM at each stage of reprocessing limit the accuracy of simple material balance accounting as a safeguards method. To be effective, a formal safeguards program must take into account all sources of measurement error yet detect any diversion of SNM. The objective of this study is to demonstrate an analytical method for assessing the accountability of selected constituent SNM. A combined discrete-continuous, time-dependent model using the GASP IV simulation language is developed to simulate mass flow, material accountability and measurement error at each stage of the reprocessing plant. The study demonstrates that the simulation method may be utilized to estimate the magnitude of SNM loss in an operating reprocessing plant which could reasonably be detected. Thus, the simulation method provides a level of confidence for effective loss/no-loss decisions. / Ph. D.

Currently in the UK, spent nuclear fuel is reprocessed using the PUREX (Plutonium Uranium Reduction Extraction) process. This process generates large amounts of aqueous nitric acid based waste which is reduced in volume by evaporation before being stored in stainless steel tanks pending eventual disposal to a repository after conversion into a solid wasteform. The corrosivity of nitric acid solutions towards these stainless steel storage tanks is strongly affected by the presence of oxidants that can form in situ if certain dissolved metals such as cerium, chromium, ruthenium and neptunium are present, which is invariably the case in nuclear reprocessing plant liquors. Such liquors are, however, subject to radiolysis leading to the formation of nitrous acid and NOx species in equilibrium with nitric acid and water. The redox chemistry of irradiated reprocessing plant liquors is therefore complex, depending on a large number of factors including acidity, nitrate ion concentration, temperature, pressure, radiation dose rate and the nature/concentration of dissolved species. High acidities, high temperatures and low dose rates favour the oxidation of species such as Ce(III). For example, when Ce(IV) forms, the corrosion rate of stainless steel is strongly increased since the reduction of Ce(IV) forms a kinetically-favoured path way. Furthermore, the presence of nitrous acid (which is radiolytically formed from nitrate/nitric acid) can act to reduce potential corrosion accelerators (e.g. Ce(IV)) to their non-oxidising valency states. These dependencies are only semi-quantitatively understood at present, hampering useful prediction of actual effects when conditions are changed. The research presented within this thesis is divided between two experimental campaigns which are interrelated by their applicability to highly active storage tank conditions; I. An investigation into the conditions which effect the radiolytic production of nitrous acid in nitric acid based solutions was undertaken. This included the quantitative measurement of the steady state concentration of nitrous acid experienced under different conditions. The conditions investigated include temperature, dose rate, gaseous headspace and liquor composition in order to elucidate which factors are of importance in estimating the concentration of nitrous acid which can be expected at the base of a highly active storage tank. The major result of this campaign was that nitrous acid data collected could be used to formulate a g-value of nitrous acid formation (which was found to be 0.71) and this value was used to calculate the nitrous acid production rate expected within a highly active storage tank which is a pre-requisite of underpinning the corrosion chemistry within. II. Investigation into the potential formation of in situ corrosion accelerators in a reprocessing liquor simulant was undertaken. For this, a bespoke experimental rig has been designed, built and operated in order to identify the valency of potential corrosion accelerators at high temperatures while closely representing the conditions expected at the base of a highly active storage tank. This included the simulation of nitric acid radiolysis by means of an appropriate nitrite addition underpinned by the radiolysis studies described above in (I). It was found that none of the conditions investigated were oxidising enough to promote the generation of Ce(IV), which is contrary to the current understanding and should be favourable with regards to the storage tank remnant life expectations. Results reported in this thesis provide insight into the corrosivity of reprocessing liquors under representative storage tank conditions at various temperatures (up to the local liquor saturation temperature) and this knowledge will help improve remnant life predictions of the highly active storage tanks on the Sellafield site.

The quantification of plutonium (Pu) in spent nuclear fuel is an increasingly important safeguards issue. There exists an estimated worldwide 980 metric tons of Pu in the nuclear fuel cycle and the majority is in spent nuclear fuel waiting for long term storage or fuel reprocessing. This study investigates utilizing the measurement of x-ray fluorescence (XRF) from the spent fuel for the quantification of its uranium (U) to Pu ratio. Pu quantification measurements at the front end of the reprocessing plant, the fuel cycle area of interest, would improve input accountability and shipper/receiver differences. XRF measurements were made on individual PWR fuel rods with varying fuel ages and final burn-ups at Oak Ridge National Laboratory (ORNL) in July 2008 and January 2009. These measurements successfully showed that it is possible to measure the Pu x-ray peak at 103.7 keV in PWR spent fuel (~1 percent Pu) using a planar HPGe detector. Prior to these measurement campaigns, the Pu peak has only been measured for fast breeder reactor fuel (~40 percent Pu). To understand the physics of the measurements, several modern physics simulations were conducted to determine the fuel isotopics, the sources of XRF in the spent fuel, and the sources of Compton continuum. Fuel transformation and decay simulations demonstrated the Pu/U measured peak ratio is directly proportional to the Pu/U content and increases linearly as burn-up increases. Spent fuel source simulations showed for 4 to 13 year old PWR fuel with burn-up ranges from 50 to 67 GWd/MTU, initial photon sources and resulting Compton and XRF interactions adequately model the spent fuel measured spectrum and background. The detector simulations also showed the contributions to the Compton continuum from strongest to weakest are as follows: the fuel, the shipping tube, the cladding, the detector can, the detector crystal and the collimator end. The detector simulations showed the relationship between the Pu/U peak ratio and fuel burn-up over predict the measured Pu/U peak but the trend is the same. In conclusion, the spent fuel simulations using modern radiation transport physics codes can model the actual spent fuel measurements but need to be benchmarked.

No description available.

Thesis (Ph.D.)--University of Tulsa, 1993. / Five colored photographs glued in book. Includes bibliographical references.

Thesis (Ph.D.)--Murdoch University, 2006. / Thesis submitted to the Division of Health Sciences. Bibliography: leaves 176-207.

Thesis (M.S.)--University of Michigan, 1976.

Thesis (Honors)--College of William and Mary, 2009. / Includes bibliographical references (leaves 21-24). Also available via the World Wide Web.

Thesis (Ph.D.)-University of KwaZulu-Natal, Durban, 2009. / Full text also available online. Scroll down for electronic link.