There has been a strong international interest in using pyroprocessing to close the fast nuclear reactor fuel cycle and reprocess spent fuel efficiently. To commercialize pyroprocessing, safeguards technologies are required to be developed. In this research, the use of Self-Interrogation Neutron Resonance Densitometry (SINRD) has been investigated as a method to safeguard the process and more precisely quantify the 239Pu content of pyroprocessing materials. This method uses a detector array with different filters to isolate the low-energy resonance in 239Pu neutron fission cross section. The relative response of the different detectors allows for the quantification of the amount of 239Pu in the pyroprocessing materials. The Monte-Carlo N-Particle (MCNP) code was used to design a prototype SINRD instrument. This instrument is composed of a neutron source pod and a SINRD detector pod. Experimental measurements were also performed to validate the MCNP model of the instrument. Based on the results from simulations and experiments, it has been concluded that the MCNP model accurately represents the physics of the experiment. In addition, different SINRD signatures were compared to identify which of them are usable to determine the fissile isotope content. Comparison of different signatures allowed for reduction in the uncertainty of the 239Pu mass estimate. Using these signatures, the SINRD instrument was shown to be able to quantify the 239Pu content of unknown pyroprocessing materials suitable for safeguards usage.

This study focuses on techniques that can be used to fuel next generation reactors. The first two studies are new techniques for recycling used nuclear fuel (UNF) and the third is a method of separating uranium (U) from lithium fluoride (LiF) and thorium fluoride (ThF4) salt also known as FLiTh for a thorium (Th) fuel cycle. The first technique proposed for UNF recycling was to use the cladding as an anode to oxidize the zircaloy and dissolve it into a LiF, sodium fluoride (NaF), zirconium fluoride (ZrF4) salt. Zirconium (Zr) was also reduced and deposited on a tungsten (W) cathode at the same time transporting the Zr through the salt. As commercial zircaloy would be contaminated with UNF oxides, and the oxides will not oxidize as part of the electrochemical process, they would be left at the anode as the Zr is dissolved away. This means the deposited Zr, on the cathode, can be disposed of as low-level waste (LLW) or recycled back into the nuclear industry instead of being stored as high-level waste (HLW). The next technique was fluorination of UNF oxides using ZrF4. Using the same LiF-NaF-ZrF4 salt, uranium oxide (UO2), lanthanum oxide (La2O3), and yttrium oxide (Y2O3) were fluorinated into uranium fluoride (UF4), lanthanum fluoride (LaF3), and yttrium fluoride (YF3). By sampling and recording the change in concentration over time, the reaction rate of all three oxides was determined and a temperature dependent reaction rate was reported from 500°C to 650°C. A zirconium oxide (ZrO2) product layer developed on UO2, but it only slowed down the fluorination process but did not stop it. UO2 and Y2O3 fluorinated entirely but La2O3 did not. The solubility limit of LaF3 in the salt was determined to be the reason the reaction did not go to completion. The last technique was the electrochemical separation of U from FLiTh, to simulate irradiated Th that decays to protactinium (Pa). A constant, albeit small current, was used to deposit U on a W electrode without Th depositing with it. A liquid metal bismuth (Bi) electrode was used as well, and a constant current resulted in Th depositing with the U. To get just U to deposit, the current needed to be applied for a time and then no current applied for a time so the system could reach equilibrium. By cycling these two steps it was possible to get U to deposit in Bi without Th. / Doctor of Philosophy / This study focused on techniques useful to the fabrication of next generation reactor fuels. The first focus was on new techniques for recycling used nuclear fuel (UNF). Nuclear waste currently needs to be stored for hundreds of thousands of years to reach background radiotoxicity levels. If plutonium (Pu) is removed from the waste this time is limited to ten thousand years and if the other transuranics (TRU) are removed the waste only needs to be stored for 300 years to reach background radiotoxicity levels. As recycling UNF can make such a drastic difference, developing techniques for this are of utmost importance. The first technique studied was to show that the zirconium (Zr) in zircaloy cladding could be oxidized and transported through salt. This was done by applying a current between a zircaloy anode and tungsten (W) cathode, dissolving the cladding into the salt. The salt used was lithium fluoride (LiF), sodium fluoride (NaF), and zirconium fluoride (ZrF4) salt called FLiNaZr. This transported Zr through the salt and then deposited it on W. If this process was done with zircaloy contaminated with used nuclear fuel (UNF) oxides, the oxides would not dissolve into the salt as part of the process and would be left behind at the anode as Zr is transported through the salt, effectively separating the two. This alone leads to a 25% reduction in the weight of the UNF that needs to be stored. The next technique studied was converting the UNF oxides into fluorides. This was done by having it react with ZrF4 to make zirconium oxide (ZrO2) and UNF fluorides. The oxides studied here were uranium oxide (UO2), yttrium oxide (Y2O3), and lanthanum oxide (La2O3). UO2 and Y2O3 reacted until no material was left but La2O3 did not. This was due to lanthanum fluoride (LaF3) having a solubility limit in the salt that made it impossible for more to be made and stopping the reacting. The reaction rate for each oxide was found and the order of the reaction rates was Y2O3>UO2>La2O3. This process was a success and should be studied more to ensure it will work with all oxides found in UNF. The last technique studied was electrochemically separating uranium (U) from lithium fluoride and thorium fluoride (ThF4) salt. Thorium (Th) is another nuclear material, and while it cannot fission in a reactor it can be turned into an isotope of U, U-233, that can. To do this Th must be irradiated so it turns into protactinium (Pa) which can then be separated from the salt. In this study U was a surrogate for Pa as it is too radioactive to handle in this lab. First, an inert W electrode was used to deposit U metal, and once it was successful a liquid metal bismuth (Bi) electrode was used. A small constant current was able to deposit U on W without co-deposition of Th. For a Bi electrode, an alternating time of applying current and then letting the system rest was needed to deposit U without co-deposition of Th.

Pyroprocessing, via electrorefining in a molten salt bath, is a promising treatment route for spent nuclear fuel reprocessing. In order to implement such a technology and ensure its safe operation it is vital to develop on-line techniques to understand and monitor the molten salt and its contents. These tools are technically challenging because of the high temperatures and corrosive environment experienced in molten salt media. Electrochemical, spectroscopic and spectroelectrochemical methods were developed and used to study actinide and fission product behaviour in molten LiCl-KCl eutectic. A spectroscopic furnace was designed and supporting methodology developed in order to allow the acquisition of reproducible quantitative data. The apparatus monitored the precipitation of NdCl3 by the addition of Li2CO3 and PrCl3 by the addition of Li2O in LiCl-KCl eutectic. The precipitates formed were identified as the respective LnOCl. In order to probe actinide behaviour in this hygroscopic medium, dry actinides chlorides were synthesised. The oxidation of uranium metal by BiCl3 in LiCl-KCl eutectic yielded UCl3 while neptunium and plutonium were prepared as Cs2AnCl6 via precipitation in concentrated aqueous HCl by addition of CsCl. The molar extinction coefficients for U(III), U(IV), Np(IV) and Pu(III) were obtained in LiCl-KCl eutectic at 450 áμ’C. The study of the Np(IV)/Np(III) couple via spectroelectrochemical techniques, enabled the determination of the Np(III) molar extinction coefficients. Uranium was studied in LiCl-KCl eutectic using square wave voltammetry, cyclic voltammetry and chronoabsorptometry. The electrochemical techniques benchmarked the results obtained by spectroelectrochemistry. The results from the different techniques were compared to and explained by determining the Gibbs energy and activation energy of U(III) and U(IV). It was concluded that all the mentioned techniques are suitable for the study of high temperature molten chlorides. Because of its capacity to gather numerous data parameters while minimising the number of experiments required and the quantity of material needed, spectroelectrochemical methods were highlighted as the most promising technique for future studies of radionuclides in high temperature melts.

The primary objective of this research is to develop an applied technology and provide an assessment for remotely measuring and analyzing the real time or near real time concentrations of used nuclear fuel (UNF) elements in electrorefiners (ER). Here, Laser-Induced Breakdown Spectroscopy (LIBS) in UNF pyroprocessing facilities was investigated. LIBS is an elemental analysis method, which is based on the emission from plasma generated by focusing a laser beam into the medium. This technology has been reported to be applicable in solids, liquids (includes molten metals), and gases for detecting elements of special nuclear materials. The advantages of applying the technology for pyroprocessing facilities are: (i) Rapid real-time elemental analysis; (ii) Direct detection of elements and impurities in the system with low limits of detection (LOD); and (iii) Little to no sample preparation is required. One important challenge to overcome is achieving reproducible spectral data over time while being able to accurately quantify fission products, rare earth elements, and actinides in the molten salt. Another important challenge is related to the accessibility of molten salt, which is heated in a heavily insulated, remotely operated furnace in a high radiation environment within an argon gas atmosphere. This dissertation aims to address these challenges and approaches in the following phases with their highlighted outcomes: 1. Aerosol-LIBS system design and aqueous testing: An aerosol-LIBS system was designed around a Collison nebulizer and tested using deionized water with Ce, Gd, and Nd concentrations from 100 ppm to 10,000 ppm. The average %RSD values between the sample repetitions were 4.4% and 3.8% for the Ce and Gd lines, respectively. The univariate calibration curve for Ce using the peak intensities of the Ce 418.660 nm line was recommended and had an R2 value, LOD, and RMSECV of 0.994, 189 ppm, and 390 ppm, respectively. The recommended Gd calibration curve was generated using the peak areas of the Gd 409.861 nm line and had an R2, LOD, and RMSECV of 0.992, 316 ppm, and 421 ppm, respectively. The partial least squares (PLS) calibration curves yielded similar results with RMSECV of 406 ppm and 417 ppm for the Ce and Gd curves, respectively. 2. High temperature aerosol-LIBS system design and CeCl3 testing: The aerosol-LIBS system was transitioned to a high temperature and used to measure Ce in molten LiCl-KCl salt within a glovebox environment. The concentration range studied was from 0.1 wt% to 5 wt% Ce. Normalization was necessary due to signal degradation over time; however, with the normalization the %RSD values averaged 5% for the mid and upper concentrations studied. The best univariate calibration curve was generated using the peak areas of the Ce 418.660 nm line. The LOD for this line was 148 ppm with the RMSECV of 647 ppm. The PLS calibration curve was made using 7 latent variables (LV) and resulting in the RMSECV of 622 ppm. The LOD value was below the expected rare earth concentration within the ER. 3. Aerosol-LIBS testing using UCl3: Samples containing UCl3 with concentrations ranging from 0.3 wt% to 5 wt% were measured. The spectral response in this range was linear. The best univariate calibration curves were generated using the peak areas of the U 367.01 nm line and had an R2 value of 0.9917. Here, the LOD was 647 ppm and the RMSECV was 2,290 ppm. The PLS model was substantially better with a RMSECV of 1,110 ppm. The LOD found here is below the expected U concentrations in the ER. The successful completion of this study has demonstrated the feasibility of using an aerosol-LIBS analytical technique to measure rare earth elements and actinides in the pyroprocessing salt.

The presence of group I/II fission products (Cs-137, Sr-90, and Ba-137) within molten salt nuclear processes degrades operational efficiencies by contributing to increased radiation levels in the case of material handling processes or to loss of criticality in the case of a reactor. While methods such as zone freezing and ion exchange have been proven for the separation of these fission products in LiCl-KCl salts, they require extra equipment and processing steps. Addition of a liquid metal electrode to molten salt media, such as the electrorefiner of a pyroprocessing scheme or the salt cleaning stage of a molten salt fast reactor, offers the possibility to selectively extract group I/II fission products with minimal extra equipment. This work has studied the fundamental electrochemical behaviors of Cs, Sr, Ba, and Ce in LiCl-KCl at a liquid Bi cathode and produced values for their mass transport and kinetics properties. These have allowed for an assessment of a Bi cathode for electrochemical separations in a pyrochemical processing scheme.

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Pyroprocessing is an electrochemical method for recovering actinides from used nuclear fuel and recycling them into fresh nuclear fuel. It is posited herein that proposed safeguards approaches on pyroprocessing for nuclear material control and accountability face several challenges due to the unproven plutonium-curium inseparability argument and the limitations of neutron counters. Thus, the Hybrid K-Edge Densitometer is currently being investigated as an assay tool for the measurement of pyroprocessing materials in order to perform effective safeguards. This work details the development of a computational model created using the Monte Carlo N-Particle code to reproduce HKED assay of samples expected from the pyroprocesses. The model incorporates detailed geometrical dimensions of the Oak Ridge National Laboratory HKED system, realistic detector pulse height spectral responses, optimum computational efficiency, and optimization capabilities. The model has been validated on experimental data representative of samples from traditional reprocessing solutions and then extended to the sample matrices and actinide concentrations of pyroprocessing. Data analysis algorithms were created in order to account for unsimulated spectral characteristics and correct inaccuracies in the simulated results. The realistic assay results obtained with the model have provided insight into the extension of the HKED technique to pyroprocessing safeguards and reduced the calibration and validation efforts in support of that design study. Application of the model has allowed for a detailed determination of the volume of the sample being actively irradiated as well as provided a basis for determining the matrix effects from the pyroprocessing salts on the HKED assay spectra.

Understanding the characteristics of special nuclear materials in LiCl-KCl eutectic salt is extremely important in terms of effective system operation and material accountability for safeguarding pyroprocessing technology. By considering that uranium (U) is the most abundant and important element in the used nuclear fuel, measurements and analyses of U properties were performed in LiCl-KCl eutectic salt. Therefore, the electrochemical techniques such as cyclic voltammetry (CV), open circuit potential (OCP), Tafel, linear polarization (LP), and electrochemical impedance spectroscopy (EIS) were conducted under different experimental conditions to explore the electrochemical, thermodynamic, and kinetic properties of U in LiCl-KCl eutectic. The ultimate goal of this study was to develop proper methodologies for measuring and analyzing the exchange current density (i0) of U3+/U reaction, which has not been fully studied and understood in literature. In the preliminary study, cerium (Ce) was selected as a surrogate material for uranium and its behavior was being explored with the developments of experimental methods. CV was performed to evaluate Ce properties such as the diffusion coefficients (D), apparent standard reduction potential (E0*), Gibbs free energy (DG), and activity coefficient (g). In addition, EIS methods were adapted and specific experimental procedures were established for the proper i0 measurements providing repeatable and reproducible data sets. The i0 values for Ce3+/Ce pair were ranging from 0.0076 A cm-2 to 0.016 A cm-2, depending on the experimental conditions. These preliminary results give insight in developing the experimental setups and methods to evaluate the properties of U in LiCl-KCl. Plus, Ce is one of the lanthanide (Ln) fission products in electrorefiner (ER) system; therefore, the resulting data values yield useful information of the fundamental behaviors of Ln elements in the system. Based on these developed methodologies, the experimental designs and routines were established to explore the main properties (e.g., D, E0*, etc.) of UCl3 in LiCl-KCl eutectic salt under different concentrations (0.5 wt% to 4 wt% UCl3) and temperatures (723 K to 798 K). Specially, the i0 values of U3+/U were evaluated via EIS, LP, Tafel, and CV methods. All i0 values had linear trends with the change of concentration and temperature; however, these values measured by LP, Tafel, and CV methods were greatly influenced by the change in electrode surface area. Overall, the i0 values agreed within 33% relative error range with the EIS method being the most consistent and accurate in comparison to reported literature values. The measured values of i0 were ranging from 0.0054 A cm-2 to 0.102 A cm-2. Therefore, an extremely reliable database for i0 was provided and it is feasible to anticipate the i0 kinetics in other experimental conditions by using the provided equation models. Furthermore, GdCl3 was added to the LiCl-KCl-UCl3 system to explore the effects of other elements on the U properties such as the diffusion coefficients, thermodynamic properties, and i0 kinetics. The diffusion coefficient was generally decreased by 12 ~ 35% with addition of GdCl3 in LiCl-KCl-UCl3; however, the apparent standard potentials and exchange current density follow the same trends with data obtained without GdCl3 additions. Hence, the results indicate that the thermodynamic and kinetic values for U3+/U reaction in LiCl-KCl eutectic salt are not greatly influenced by the presence of GdCl3.

Future nuclear fission reactors (GEN IV) are designed to include fast breeder reactor technologies, which can accept transuranics (elements heavier than uranium) as fuel. This has the potential of being more fuel efficient but requires the closing of the nuclear fuel cycle: full recycling of existing and newly generated nuclear waste to extract uranium and transuranic elements which can be reused as fuel. In the UK a system being investigated is electrochemical pyroprocessing which uses molten LiCl-KCl eutectic (LKE), which aims to recover uranium by electrodeposition on an inert (steel) electrode and the transuranics by electrodeposition as alloys with an active metal electrode (bismuth, cadmium or aluminium). Of the three active metal candidates, aluminium has the best separation efficiency of actinides and lanthanides, which is important as lanthanides are neutron poisons and so are not to be extracted. The development of pyroprocessing requires fundamental understandings of electrochemical alloy formation, as well as on-line monitoring tools to ensure the reprocessing occurs safely and efficiently. To that end, this thesis investigates cerium-aluminium alloying (a non-radioactive model system for plutonium-aluminium) on macro- and microelectrodes to understand the limiting factors during the alloying reaction at each electrode scale and also the circumstances under which the Ce3+ concentration can be reliably determined for on-line monitoring. On a bulk aluminium macroelectrode one cerium-aluminium alloying reaction was observed. This reaction was kinetically limited by the phase change from cerium insertion into the aluminium, and resulted in lattice expansion and progressive roughening of the electrode surface. These factors made it difficult to reliably calculate the Ce3+ concentration. Li+ from the solution was also able to reduce and form alloys with aluminium, approximately 0.3 V more negative than the first cerium-aluminium alloying peak. Since lithium atoms are smaller than cerium, and there is an abundance of Li+ in the salt, lithium-aluminium alloy was found to form preferentially to cerium-aluminium alloy at these more negative potentials. By co-depositing Al3+ and Ce3+ together on a tungsten electrode which is inert under these conditions (it does not alloy), the kinetic barrier to alloy formation by cerium insertion was decreased, which is beneficial to studying the thermodynamics of alloying. Studies of pure aluminium plating and pure cerium plating showed each individual reaction was diffusion limited, with an increased contribution of convection to the mass transport at slow scan rates. Co-deposition on macroelectrodes with a low ratio of [CeCl3]:[AlCl3] showed only one cerium-aluminium alloying peak. The co-deposition currents, and ratio of oxidation peaks charges, showed that co-deposition was occurring with both species under diffusion control, resulting in an amorphous alloy with a Ce:Al ratio that smoothly varied with the [CeCl3]:[AlCl3] ratio. This was in contrast to the alloying behaviour of cerium with liquid bismuth, in which co-deposition occurred at specific ratios determined by the crystal phases that could be formed at the applied potentials, with higher co-deposition ratios being achieved at more negative potentials. Co-deposition on macroelectrodes with a high ratio of [CeCl 3]:[AlCl3] could result in up to five cerium-aluminium alloy peaks, corresponding to all five CexAly crystalline phases predicted by the phase diagram. This phase change from amorphous to crystalline was promoted by the high Ce:Al ratio in the amorphous alloy resulting from the high [CeCl3]:[AlCl3] ratio and by plating pure cerium on the surface, which could then insert into the alloy. Charge analysis of these peaks confirmed the expected stoichiometries of the crystal phase from these in-situ measurements which is important for rapid analysis, whereas all previous literature has relied on ex-situ techniques which cooled the alloy, possibly changing its composition and structure. In all circumstances of alloy formation on macroelectrodes, the rate of reduction of Ce3+ was time dependent and sensitive to convection. This significantly complicated analysis of the electrochemical signal, making it very difficult to reliably calculate the concentration of Ce3+, which is required for on-line monitoring. Co-deposition on in-house microfabricated tungsten microelectrodes resulted in steady state currents for both pure aluminium deposition and cerium-aluminium co-deposition (up to the beginning of lithium-aluminium alloying). Thus, unlike on macroelectrodes, the deposition rate occurred at the flux ratio of each species from solution and only one oxidation peak was observed corresponding to the amorphous cerium-aluminium phase, even at high [CeCl3]:[AlCl3] ratios. The steady state alloying current meant that calculating the Ce3+ concentration was relatively simple from co-deposition on microelectrodes. Co-deposition was highly beneficial for studying alloying, however to avoid the addition of Al3+ to the molten salt, in-house microfabricated thin film aluminium microelectrodes were also used to study alloying. Alloying on microfabricated thin film aluminium microelectrodes was hampered by the formation of a native aluminium oxide layer, which prevented cerium insertion into the aluminium. The oxide layer could be disrupted by reduction of lithium, which showed steady state currents (albeit with significant capacitance) could be achieved for alloying by cerium insertion. However, the full surface area of the microelectrode could not be attained and all microelectrodes lost their aluminium layer after multiple lithiation/de-lithiation cycles. These devices need further development to overcome the oxide layer, or prevent its formation, in order to study alloying in greater detail with aluminium microelectrodes to fully realise their advantages for sensing and monitoring in pyroprocessing.