<p>The Encyclopedia of Life Support Systems defines sustainability or industrial ecology as ?the wise use of resources through critical attention to policy, social, economic, technological, and ecological management of natural and human engineered capital so as to promote innovations that assure a higher degree of human needs fulfilment, or life support, across all regions of the world, while at the same time ensuring intergenerational equity? (Encyclopedia of Life Support Systems 1998). Developing and integrating sustainable energy systems to meet growing energy demands is a daunting task. Although the technology to utilize renewable energies is well understood, there are limited locations which are ideally suited for renewable energy development. Even in areas with significant wind or solar availability, backup or redundant energy supplies are still required during periods of low renewable generation. This is precisely why it would be difficult to make the switch directly from fossil fuel to renewable energy generation. A transition period in which a base-load generation supports renewables is required, and nuclear energy suits this need well with its limited life cycle emissions and fuel price stability. Sustainability is achieved by balancing environmental, economic, and social considerations, such that energy is produced without detriment to future generations through loss of resources, harm to the environment, etcetera. In essence, the goal is to provide future generations with the same opportunities to produce energy that the current generation has. This research explores sustainability metrics as they apply to a small modular reactor (SMR)-hydrogen production plant coupled with wind energy and storage technologies to develop a new quantitative sustainability metric, the Sustainability Efficiency Factor (SEF), for comparison of energy systems. The SEF incorporates the three fundamental aspects of sustainability and provides SMR or nuclear hybrid energy system (NHES) reference case studies to (1) introduce sustainability metrics, such as life cycle assessment, (2) demonstrate the methods behind exergy and exergoeconomic analyses, (3) provide an economic analysis of the potential for SMR development from first-of-a-kind (FOAK) to nth-of-a-kind (NOAK), thereby illustrating possible cost reductions and deployment flexibility for SMRs over large conventional nuclear reactors, (4) assess the competitive potential for incorporation of storage and hydrogen production in NHES and in regulated and deregulated electricity markets, (5) compare an SMR-hydrogen production plant to a natural gas steam methane reforming plant using the SEF, and (6) identify and review the social considerations which would support future nuclear development domestically and abroad, such as public and political/regulatory needs and challenges. The Global Warming Potential (GWP) for the SMR (300 MWth)-wind (60 MWe)-high temperature steam electrolysis (200 tons Hydrogen per day) system was calculated as approximately 874 g CO2-equivalent as part of the life cycle assessment. This is 92.6% less than the GWP estimated for steam methane reforming production of hydrogen by Spath and Mann. The unit exergetic and exergoeconomic costs were determined for each flow within the NHES system as part of the exergy/exergoeconomic cost analyses. The unit exergetic cost is lower for components yielding more meaningful work like the one exiting the SMR with a unit exergetic cost of 1.075 MW/MW. In comparison, the flow exiting the turbine has a very high unit exergetic cost of 15.31, as most of the useful work was already removed through the turning of the generator/compressor shaft. In a similar manner, the high unit exergoeconomic cost of $12.45/MW*sec is observed for the return flow to the reactors, because there is very little exergy present. The first and second law efficiencies and the exergoeconomic factors were also determined over several cases. For the first or base SMR case, first and second law efficiencies of 81.5% and 93.3% were observed respectively. With an increase in reactor outlet temperature of only 20?C, both the SMR efficiencies increased, while the exergoeconomic factor decreased by 0.2%. As part of the SMR economic analysis, specific capital and total capital investment costs (TCIC) were determined in addition to conditional effects on the net present value (NPV), levelized cost of electricity (LCOE), and payback periods. For a 1260 MWe FOAK multi-module SMR site with 7 modules, the specific capital costs were 27-38% higher than that of a 1260 MWe single large reactor site. A NOAK site, on the other hand, may be 19% lower to 18% higher than the large reactor site, demonstrating that it may break even or be even more economical in average or favorable market conditions. The NOAK TCIC for single and multi-module SMR sites were determined to be $914-$1,230 million and $660-$967 million per module, respectively, reflecting the substantial savings incurred with sites designed for and deployed with multiple modules. For the same NOAK 7-unit multi-module site, the LCOE was calculated as $67-$84/MWh, which is slightly less than that of the conventional large reactor LCOE of $89/MWh with a weighted average cost of capital of 10%, a 50%-50% share of debt and equity, and a corporate tax rate of 35%. The payback period for the SMR site, however, is 4 years longer. Construction delays were also analyzed to compare the SMR and large reactor sites, demonstrating the SMR NPV and LCOE are less sensitive to delays. For a 3 year delay, the SMR NPV decreased by 22%, while the large reactor NPV decreased by 34.1%. Similarly the SMR and large reactor LCOEs increased by 7.8% and 8.1%, respectively. An NHES case with hydrogen production and storage was performed, illustrating how the profit share of revenue is improved with the addition of hydrogen production. Although the costs are increased with the addition, 78% of the hydrogen revenue is profit, while only 50% of the electricity generation revenue is profit. A second NHES case study was analyzed to assess the NPV, LCOE, and payback differences in deregulated and regulated electricity markets. For a 60 year lifetime, Case C (with nuclear, wind, and hydrogen production) is economical in the deregulated market with an NPV of ~$66.3 million and a payback period of 10 years, but not in the regulated one with an NPV of approximately -$115.3 million and a payback period of 11 years. With either market type, the plants levelized costs remain $82.82/MWh, which is still reasonable with respect to prior LCOE values determined for SMR and large reactor sites. Utilizing all the methodology and results obtained and presented in this thesis, the SEF may be calculated. The NHES SEF was determined to be 18.3% higher than that of natural gas steam methane reforming, illustrating a higher level of sustainability. The SEF quantitatively uses the exergoeconomic cost and irreversibilities obtained from the exergy analysis, the GWP obtained from the life cycle assessment and costs/fees associated with emissions and pollutants, and relevant economic data obtained from an economic analysis. This reflects the environmental, socio-political, and economic pillars of sustainability.

To evaluate the release of some important radionuclides from cement waste forms, a variety of studies have been completed. Leach rates and equilibrium concentrations of some complexing agents and radionuclides were measured, and the effect of a simulated waste form on these values was determined. Complexing agent concentrations were evaluated. Tests in deionized water showed ethylenediaminetetraacetic acid (EDTA) was leached from a simulated waste form more rapidly than nitrilotriacetic acid (NTA) which was released more rapidly than citric acid. Equilibrium concentrations were measured in deionized water, saturated sodium chloride, 24% magnesium chloride and Q-brine which is a simulated groundwater based on magnesium chloride. At equilibrium, the EDTA release fraction was independent of the amount of EDTA in the cement. The citric acid concentration was constant at less than 10⁻⁵ M. Finally, NTA concentrations and release fractions were both dependent on the cement loading. To evaluate the effect of complexing agents on some radionuclides, leach rates of cesium and cobalt and equilibrium concentrations of cobalt and uranium were studied. The effects of EDTA, NTA and citric acid on release rates of cesium and cobalt were studied, and the effects of several complexing agents on the equilibrium concentrations of cobalt and uranium were examined. Complexing agents had no effect on cesium leach rates. In water and in sodium chloride, cobalt and uranium concentrations in the presence of cement did not vary with any complexing agent. In Q-brine in the presence of cement, tartaric acid, EDTA and NTA increased the soluble cobalt concentration two orders of magnitude, and citric acid had a much smaller effect. Uranium concentrations in Q-brine were increased in EDTA, tartaric acid and citric acid when cement was in the system.

A new development and practical application of a mathematical dynamic modeling for simulating normal and accidental transient analysis for the boiling water reactor system is presented in this dissertation. The mathematical dynamic modeling represents a new technology based on a moving boundary concept. The mathematical model developed for fluid flows is based on a set of the four equation mixture model, one-dimensional, single channel with a drift flux model in the two-phase flow regime. The four conservation equations used in the mathematical model formulation include the vapor phase mass equation, the liquid phase mass equation, the mixture energy equation, and the one-dimensional mixture momentum equation for the boiling channel. The formulation of the core thermal-hydraulic model utilizes a transient moving boundary technique which tracks the movements of the phase change and boiling transition boundaries. Such a moving boundary model has been developed to allow a smooth representation of the boiling boundary movement based on empirical heat transfer correlations and the local thermal-hydraulic conditions of the coolant flow along fuel pin channels. The mathematical models have been implemented to accommodate three-dimensional reactor kinetics, with detailed thermal conduction in fuel elements. Also, an accurate minimum departure from nucleate boiling ratio (MDNBR) boundary is predicted during transients. Several test calculations were performed to assess the accuracy and applicability of the moving boundary model. Comparison between the calculated results and the experimental data are favorable. Overall system studies show that some thermal margin is gained using the transient MDNBR approach vs the traditional quasi-static methodology. The model predicts accurate void fraction profiles for kinetic feedback and boiling stability analysis for the BWR. The moving boundary formulation and improved numerical solution scheme are an efficient and suitable tool which can be useful for realistic simulation of degraded nuclear power plant transients.

Wall plasmas, plasmas in contact with a solid wall, are studied in two basic cases. The first is a linear theta-pinch. In this geometry the motion of the plasma along the field lines and the transfer of heat along the field lines is investigated using a transient hydrodynamic heat flow model and a steady state heat flow model. Good agreement is found with experimental results on the Scylla IV-P linear theta pinch. Results of the stationary model are used to estimate linear fusion reactor lengths. Results are that linear fusion reactors would be shorter for low atomic number (Z) end walls, but have higher linear power levels (power per unit length) for high Z end walls. The second case is a stationary solution of the structure of a plasma in contact with a diverter or limiter in a tokamak. This geometry is investigated to determine relationships between plasma density, temperature and the velocity of plasma flow toward the diverter. Solutions are found and sensitivities are presented.

Fissile materials in the form of aqueous homogeneous solutions are used during the chemical processing of nuclear fuel. In this form there exists the possibility of an accidental criticality of the solution. To determine the consequences of such accidents, computer models have been developed to simulate nuclear excursions. A one-region model and a multi-region model have been developed to simulate both power and pressure pulses. These models include a new radiolytic gas production model that tracks the number of radiolytic gas bubbles produced during an excursion. Also, an equation of state, which accounts for the production of inertial pressure due to a "lag" in thermal expansion and the creation of radiolytic gas bubbles, has been developed for both models. The multi-region model can account for the spatial distribution of the nuclear energy deposited in the solution and both axial and radial acceleration of the fuel material caused by the production of inertial pressure. Predicted power and pressure pulses have been compared with experimental data from the KEWB, CRAC, and SILENE experiments. The computer models have been very successful in predicting the magnitude of both power and pressure pulses. Also, the multi-region model has provided new information on the spatial distribution of solution parameters during an excursion.

An integrated model that estimates loop flow rate, heat removal, and stability parameters for the General Electric Simplified Boiling Water Reactor SBWR was developed. The three parameters above used to be calculated individually each by a separate code. The initial approach in loop thermal hydraulic modelling was the steady state solution of the SBWR loop mass, energy, and momentum equations. The power-to-flow map obtained proved to be quite comparable with the Natural Circulation in Boiling Water Reactor (NATBWR) code developed by EPRI, in addition to that of General Electric. At low power levels buoyancy forces are the controlling factor in determining the loop flow rate, while at high power levels two-phase friction losses become the dominating one. Evaluation criteria necessary for comparing different loop geometries performance have been the "minimum critical heat flux ratio (MCHFR)" and the "decay ratio." The predicted flow, from the DFM, at different power levels was used later in a parametric study to answer an important question of which combination of core and riser heights are to be selected that meets both the stability and critical power ratio limits. By modelling bubble time delay through riser in the loop momentum equation, a loop damping coefficient as a measure of loop stability, with higher damping meaning a more stable loop was calculated. Results indicated that during normal operation the SBWR loop is pretty damped. Finally, a detailed code that consists mainly of a fuel pin model, reactor point kinetics for the time dependent reactor normalized power with one group of delayed neutrons, and coolant channel mass, energy, and momentum equations is considered. Reactivity feedbacks from voids and fuel temperature, (Doppler effect), were considered. The loop momentum equation was modified to account for bubble time delay in the riser. After a small perturbation in reactivity, fuel temperature, core average void, and loop flow rate were shown to reach equilibrium values after a period of time equivalent to the transit time of the bubble through the riser. Results from this code matched that of the SBWR safety analysis report.

The capabilities of the coupled charged/neutral particle transport S(N) code SMARTEPANTS (Simulating Many Accumulative Rutherford Trajectories Electron Photon and Neutral Transport Solver) have been extended from x-y-z geometry to x-y-z geometry with embedded cylinders. A new method called the super-cell algorithm was applied to accommodate cylindrical shapes using a rectangular mesh. The super-cell is defined as a rectangular mesh cell containing one or more material interfaces. Each material region within a super-cell constitutes a sub-cell. To model cylindrical shapes, curved sub-cell interfaces were used. The critical aspect of the super-cell method was to determine the angular fluxes in a sub-cell within the super-cell. To do this, the super-cells were divided into two major categories, Type-1 and Type-2. The cylinder's radius compared to the super-cell's mesh size was used as a basis to distinguish the super-cell's type. Each type was divided into several sub-cases depending on the direction cosine of the angular flux when entering the super-cell. The super-cell method was integrated into SMARTEPANTS and then used to calculate the energy deposition for a variety of test problems to check the method's sensitivity to its parameters. For a block of galium-arsenide (Ga-As) with an embedded gold cylinder, it was found that an S₈ quadrature set with a five energy groups is both time efficient and yields satisfactory results. The effect of cylinder radius compared to the mesh size in Type-2 super-cells was found to be minimum for an optimum mesh size. Several benchmark problems were performed to compare the super-cells results with coupled electron/photon Monte Carlo code (ITS). The total energy deposition in the peak energy cell was selected to facilitate the comparison. Peak energy cell is the cell with the maximum energy deposition. For an isotropic electron source in a Ga-As block embedded with Type-1 and Type-2 gold cylinders the results were within 3% and 6% respectively and SMARTEPANTS results in the non-super-cells were more symmetric than Monte Carlo. Super-cell also demonstrated better computer efficiency both in CPU time and memory when compared with the Monte Carlo method on the same machine.

No description available.

<p> Air bubble injection and subcooled flow boiling experiments have been performed to investigate the liquid flow field and bubble nucleation, growth, and departure, in part to contribute to the DOE Nuclear HUB project, Consortium for Advanced Simulation of Light Water Reactors (CASL). The main objective was to obtain quantitative data and compartmentalize the many different interconnected aspects of the boiling process &mdash; from the channel geometry, to liquid and gas interactions, to underlying heat transfer mechanisms. </p><p> The air bubble injection experiments were performed in annular and rectangular geometries and yielded data on bubble formation and departure from a small hole on the inner tube surface, subsequent motion and deformation of the detached bubbles, and interactions with laminar or turbulent water flow. Instantaneous and ensemble- average liquid velocity profiles have been obtained using a Particle Image Velocimetry technique and a high speed video camera. Reynolds numbers for these works ranged from 1,300 to 7,700.</p><p> Boiling experiments have been performed with subcooled water at atmospheric pres- sure in the same annular channel geometry as the air injection experiments. A second flow loop with a slightly larger annular channel was constructed to perform further boiling experiments at elevated pressures up to 10 bar. High speed video and PIV measurements of turbulent velocity profiles in the presence of small vapor bubbles on the heated rod are presented. The liquid Reynolds number for this set of experiments ranged from 5,460 to 86,000. It was observed that as the vapor bubbles are very small compared to the injected air bubbles, further experiments were performed using a microscopic objective to obtain higher spatial resolution for velocity fields near the heated wall. Multiple correlations for the bubble liftoff diameter, liftoff time and bub- ble history number were evaluated against a number of experimental datasets from previous works, resulting in a new proposed correlations that account for fluid prop- erties that vary with pressure, heat flux, and variations in geometry.</p>

<p>The production of Molybdenum-99 (99Mo) is critical to the field of nuclear medicine, where it is utilized in roughly 80% of all nuclear imaging procedures. In October of 2016, the National Research Universal (NRU) reactor in Canada, which historically had the highest 99Mo production capability worldwide, ceased routine production and will be permanently shut down in 2018. This loss of capacity has led to widespread concern over the ability of the 99Mo supply chain and to meet demand. There is significant disagreement among analyses from trade groups, governments, and other researchers, predicting everything from no significant impact to major worldwide shortages. Using Bayesian networks, this research focused on modeling the 99Mo supply chain to quantify how a disrupting event, such as the unscheduled downtime of a reactor, will impact the global supply. This not only includes quantifying the probability of a shortage occurring, but also identifying which nodes in the supply chain introduce the most risk to better inform decision makers on where future facilities or other risk mitigation techniques should be applied.