Thesis (Ph. D.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2009. / Cataloged from PDF version of thesis. / Includes bibliographical references (p. 292-301). / A robust and reliable code to model the irradiation behavior of metal and oxide fuels in sodium cooled fast reactors is developed. Modeling capability was enhanced by adopting a non-empirical mechanistic approach to the extent possible, so that to increase the ability to extrapolate the existing database with a reasonable accuracy. Computational models to analyze in-reactor behavior of U-Zr and U-Pu-Zr metallic alloy fuel pins and U0 2-PuO 2 mixed oxide fuel pins have been developed and implemented into a new code, the Fuel Engineering and Structural analysis Tool (FEAST). FEAST has several modules working in coupled form with an explicit numerical algorithm. These modules describe (1) Fission Gas Release and Swelling, (2) Fuel Chemistry and Restructuring, (3) Temperature Distribution, (4) Fuel Clad Chemical Interaction, (5) Fuel and Clad Mechanical Analysis and (6) Transient Creep- Fracture Model for the clad. Given the fuel pin geometry, composition and irradiation history, FEAST can analyze fuel and cladding thermo-mechanical behavior at both steady state and design-basis accident scenarios. FEAST was written in FORTRAN-90 program language. The FEAST-METAL code mechanical analysis module implements the old Argonne National Laboratory (ANL)'s LIFE code algorithm. / (cont.) Fission gas release and swelling are modeled with the Korean GRSIS algorithm, which is based on detailed tracking of the fission gas bubbles within the metal fuel. Migration of the fuel constituents is modeled by means of the thermo-transport theory. Fuel Clad Chemical Interaction (FCCI) models were developed for steady-state and transient situations, based on precipitation kinetics. A transient creep fracture model for the clad, based on the constrained diffusional cavity growth model, was adopted. FEAST-METAL has been benchmarked against available EBR-II database for (steady state) and furnace tests (transients). The results show that the code is able to predict important phenomena such as cladding strain, fission gas release, clad wastage, clad failure time and axial fuel slug deformation, satisfactorily. A similar code for oxide fuels, FEAST-OXIDE, was also developed. It adopts the OGRES model to describe fission gas release and swelling. However, the original OGRES model has been extended to include the effects of Joint Oxide Gain (JOG) formation on fission gas release and swelling. The fuel chemistry model includes diffusion models for radial actinide migration, cesium axial and radial migration, formation of the JOG, and variation of the oxygen to metal ratio. Fuel restructuring is also modeled, and includes the effects of porosity migration, irradiation-induced fuel densification and grain growth. / (cont.) The FEAST-OXIDE predictions has been compared to the available FFTF, EBR-II and JOYO databases, and the agreement between the code and data was found to be satisfactory. Both metal and oxide versions of FEAST are rather superior compared to many other fuel codes in the literature. Comparing metal and oxide versions, FEAST-OXIDE has a more sophisticated fission gas release and swelling model, which is based on vacancy flow. In addition, modeling of the chemistry module of the oxide fuel requires a much more detailed analysis to estimate its impact on the thermo-mechanical behavior with a reasonable accuracy. Finally, the melting of the oxide fuel and its effect on the thermo-mechanical performance have been modeled in case of transient scenarios. / by Aydin Karahan. / Ph.D.

Thesis (S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2012. / Cataloged from PDF version of thesis. / Includes bibliographical references. / The MIT Research Reactor (MITR) is evaluating the conversion from highly enriched uranium (HEU) to low enrichment uranium (LEU) fuel. In addition to the fuel element re-design from 15 to 18 plates per element, a reactor power upgraded from 6 MW to 7 MW is proposed in order to maintain the same reactor performance of the HEU core. Previous approaches in analyzing the impact of engineering uncertainties on thermal hydraulic limits via the use of engineering hot channel factors (EHCFs) were unable to explicitly quantify the uncertainty and confidence level in reactor parameters. The objective of this study is to develop a methodology for MITR thermal hydraulic limits analysis by statistically combining engineering uncertainties in order to eliminate unnecessary conservatism inherent in traditional analyses. This methodology was employed to analyze the Limiting Safety System Settings (LSSS) for the MITR LEU core, based on the criterion of onset of nucleate boiling (ONB). Key parameters, such as coolant channel tolerances and heat transfer coefficients, were considered as normal distributions using Oracle Crystal Ball for the LSSS evaluation. The LSSS power is determined with 99.7% confidence level. The LSSS power calculated using this new methodology is 9.1 MW, based on core outlet coolant temperature of 60 'C, and primary coolant flow rate of 1800 gpm, compared to 8.3 MW obtained from the analytical method using the EHCFs with same operating conditions. The same methodology was also used to calculate the safety limit (SL) to ensure that adequate safety margin exists between LSSS and SL. The criterion used to calculate SL is the onset of flow instability. The calculated SL is 10.6 MW, which is 1.5 MW higher than LSSS, permitting sufficient margin between LSSS and SL. / by Keng-Yen Chiang. / S.M.

Thesis: S.M., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 2017. / Cataloged from PDF version of thesis. / Includes bibliographical references (pages 145-149). / As discovered during the Molten Salt Reactor (MSR) Experiment (MSRE), selected fission products deposited on the wetted surfaces throughout the reactor. Fission products such as molybdenum and ruthenium are noble with respect to the electrochemical potential of the fluoride fuel salt and therefore remain insoluble in their elemental forms rather than becoming ionic salts. Coalescing in the primary fluid, these noble metals then migrate and eventually deposit on internal reactor surfaces. Since the bulk of these noble metal fission products are also energetically unstable they bring not only physical fouling, but heat and radiation from decay as well. The adherence forces of five of the seven principal radioactive foulants discovered during the MSRE were measured on six different potential structural materials by atomic force microscope force spectroscopy (AFM-FS). The noble metals studied were niobium, molybdenum, ruthenium, antimony and tellurium. Structural materials measured were Hastelloy-N, the primary structural metal of the MSRE; two steels, SS316L and F91; commercially pure nickel and molybdenum; and silicon carbide. MSRs operate with surfaces free of passivating corrosion layers, so the measurements were conducted on bare metal surfaces. An argon ion gun chamber was constructed for removal of the oxide layers from mechanically polished substrate metals by sputtering. A combination vacuum chamber/glove box was crafted to accept the sputter polished substrates into a dry, inert environment where the force adhesion measurements were made. Data acquired in the last phase of the study partially demonstrated the concept. The measured particle-to-substrate attractive forces found antimony and tellurium to be generally more adherent to the bare metals than niobium or molybdenum. The finding is consistent with the fouling examined in the MSRE. If held, this correlation of laboratory measurement to actual fouling may aid the reactor designer in anticipating fouling to plan for the effects. Such knowledge would inform selection of plant materials, both for operating components such as flow detectors and heat exchangers, where fission product deposition would be undesired, and for processing components such as filters and metal collection systems, where adhesion would be preferred. / by Reid S. Tanaka. / S.M.

Thesis: Ph. D., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 2014. / Cataloged from PDF version of thesis. / Includes bibliographical references (pages 172-178). / The performance of zirconium alloys in nuclear reactors is compromised by corrosion and hydrogen pickup. The thermodynamics and kinetics of these two processes are governed by the behavior of point defects in the ZrO₂ layer that grows natively on these alloys. In this thesis, we developed a general, broadly applicable framework to predict the equilibria of point defects in a metal oxide. The framework is informed by density functional theory and relies on notions of statistical mechanics. Validation was performed on the tetragonal and monoclinic phases of ZrO₂ by comparison with prior conductivity experiments. The framework was applied to four fundamental problems for understanding the corrosion and hydrogen pickup of zirconium alloys. First, by coupling the predicted concentrations of oxygen defects in tetragonal ZrO₂ with their calculated migration barriers, we determined oxygen self-diffusivity in a wide range of thermodynamic conditions spanning from the metal-oxide interface to the oxide-water interface. This facilitates future macro-scale modeling of the oxide layer growth kinetics on zirconium alloys. Second, using the computed defect equilibria of the tetragonal and monoclinic phases, we constructed a temperature-oxygen partial pressure phase diagram for ZrO₂. The diagram showed that the tetragonal phase can be stabilized below its atmospheric transition-temperature by lowering the oxygen chemical potential. This work adds a new explanation to the stabilization of the tetragonal phase at the metal-oxide interface where the oxygen partial pressure is low. Third, using the developed framework, we modeled co-doping of monoclinic ZrO₂ with hydrogen and a transition metal. Our modeling predicted a volcano-like dependence of hydrogen (proton) solubility on the first-row transition metals, which is consistent with a set of systematic experiments from the nuclear industry. We discovered that the reason behind this behavior is the ability of the transition metal to p-type-dope ZrO₂ and hence lower the chemical potential of electron. Therefore, the peak of the hydrogen solubility in monoclinic ZrO₂ also corresponds to an increased barrier for hydrogen gas evolution on the surface. This explanation opens the door to physics-based design of resistant zirconium alloys, and qualitatively consistent with the monoclinic ZrO₂. Finally, we uncovered the interplay between certain hydrogen defects and planar compressive stress which tetragonal ZrO₂ experiences on zirconium alloys. The stress enhances the abundance of these defects, while these same defects tend to relax the stress. This interplay was used to propose an oxide fracture mechanism by which hydrogen is picked up. / by Mostafa Youssef Mahmoud Youssef. / Ph. D.

Thesis: Ph. D. in Computational Nuclear Science and Engineering, Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 2018. / This electronic version was submitted by the student author. The certified thesis is available in the Institute Archives and Special Collections. / Cataloged from student-submitted PDF version of thesis. / Includes bibliographical references (pages 177-188). / A central goal in computational nuclear engineering is the high-fidelity simulation of a full nuclear reactor core by way of a general simulation method. General full core simulations can potentially reduce design and construction costs, increase reactor performance and safety, reduce the amount of nuclear waste generated, and allow for much more complex and novel designs. To date, however, the time to solution and memory requirements for a general full core high fidelity 3D simulation have rendered such calculations impractical, even using leadership class supercomputers. Reactor designers have instead relied on calibrated methods that are accurate only within a narrow design space, greatly limiting the exploration of innovative concepts. One numerical simulation approach, the Method of Characteristics (MOC), has the potential for fast and efficient performance on a variety of next generation computing systems, including CPU, GPU, and Intel Xeon Phi architectures. While 2D MOC has long been used in reactor design and engineering as an efficient simulation method for smaller problems, the transition to 3D has only begun recently, and to our knowledge no 3D MOC based codes are currently used in industry. The delay of the onset of full 3D MOC codes can be attributed to the impossibility of "naively" scaling current 2D codes into 3D due to prohibitively high memory requirements. To facilitate transition of MOC based methods to 3D, we have developed a fundamentally new computational algorithm. This new algorithm, known as The Random Ray Method (TRRM), can be viewed as a hybrid between the Monte Carlo (MC) and MOC methods. Its three largest advantages compared to MOC are that it can handle arbitrary 3D geometries, it offers extreme improvements in memory efficiency, and it allows for significant reductions in algorithmic complexity on some simulation problems. It also offers a much lower time to solution as compared to MC methods. In this thesis, we will introduce the TRRM algorithm and a parallel implementation of it known as the Advanced Random Ray Code (ARRC). Then, we will evaluate its capabilities using a series of benchmark problems and compare the results to traditional deterministic MOC methods. A full core simulation will be run to assess the performance characteristics of the algorithm at massive scale. We will also discuss the various methods to parallelize the algorithm, including domain decomposition, and will investigate the new method's scaling characteristics on two current supercomputers, the IBM Blue Gene/Q Mira and the Cray XC40 Theta. The results of these studies show that TRRM is capable of breakthrough performance and accuracy gains compared to existing methods which we demonstrate to enable general, full core 3D high-fidelity simulations that were previously out of reach. / by John Robert Tramm. / Ph. D. in Computational Nuclear Science and Engineering

Thesis (S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2013. / Cataloged from PDF version of thesis. / Includes bibliographical references (pages 176-179). / This study has extended the development of an exciting technology which promises to enable the Pb-Bi eutectic cooled reactors to operate at temperatures up to 650-700°C. This new technology is a functionally graded composite steel which resists high temperature LBE corrosion. This composite steel consists of a Fel2Cr2Si protective layer weld overlaid on a T91 steel and then drawn to fuel cladding and piping material. A series of tests and materials analysis were performed on the composite piping material. These tests / analysis included microstructural characterization, heat treatment optimization, creep and tensile testing, diffusion testing, and long term static corrosion tests. Although the composite fuel cladding was not available at the time of this study, all of the results from the piping material characterization are directly applicable to the fuel cladding material. It has been shown that the heat treated composite piping material exhibits mechanical properties in excess of the ASTM minimum standard for T91. This material also exhibits a conservative corrosion rate of< 22pm/yr in static Pb-Bi eutectic. This low corrosion rate will enable fuel cladding to have a 3.6 year lifetime and piping material a 36 year lifetime, if the static corrosion rate is equivalent to the flowing corrosion rate. This material has also been shown to have a very slow diffusion rate for chromium, with a chromium inter-diffusion zone of < 35um over the lifetime of the nuclear reactor. There still however exist several challenges to implementing this technology. The challenges include resolving the issue of cracking of the Fel2Cr2Si layer during tube drawing and increasing the high temperature stress / creep resistance of the structural T91 layer. / by Elliott Shepard Fray. / S.M.

Thesis (S.M. and S.B.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2006. / Includes bibliographical references (p. 109-111). / Quantification of human error probabilities (HEPs) for the purpose of human reliability assessment (HRA) is very complex. Because of this complexity, the state of the art includes a variety of HRA models, each with its own objectives, scope and quantification method. In addition to varying methods of quantification, each model is replete with its own terminology and categorizations, therefore making comparison across models exceedingly difficult. This paper demonstrates the capabilities and limitations of two prominent HRA models: the Electric Power Research Institute (EPRI) HRA Calculator (using the HRC/ORE and Cause Based Decision Tree methods), used widely in industry, and A Technique for Human Error Analysis (ATHEANA), developed by the US Nuclear Regulatory Commission. This demonstration includes a brief description of the two models, a comparison of what they incorporate in HEP quantification, a "translation" of terminologies, and examples of their capabilities via the Halden Task Complexity experiments. Possible ways to incorporate learning from simulator experiments, such as those at Halden, to improve the quantification methods are also addressed. The primary difference between ATHEANA and the EPRI HRA Calculator is in their objectives. EPRI's objective is to provide a method that is not overly resource intensive and can be used by a PRA analyst without significant HRA experience. Consequently, EPRI quantifies HEPs using time reliability curves (TRCs) and cause based decision trees (CBDT). ATHEANA attempts to find contexts where operators are likely to fail without recovery and quantify the associated HEP. This includes finding how operators can further degrade the plant condition while still believing their actions are correct. ATHEANA quantifies HEPs through an expert judgment elicitation process. / (cont.) ATHEANA and the EPRI Calculator are very similar in the contexts they consider in HEP calculation: both factor in the accident sequence context, performance shaping factors (PSFs), and cognitive factors into HEP calculation. However, stemming from the difference in objectives, there is a difference in how deeply into a human action each model probes. ATHEANA employs a HRA team (including a HRA expert, operations personnel and a thermo-hydraulics expert) to examine a broad set of PSFs and contexts. It also expands the accident sequences to include the consequences of a misdiagnosis beyond simple failures in implementing the procedures (what will the operator likely do next given a specific misdiagnosis?) To limit the resource burden, the EPRI Calculator is prescriptive and limits the PSFs and cognitive factors for consideration thus enhancing consistency among analysts and reducing needed resources. However, CBDT and ATHEANA have the same approach to evaluating the cognitive context. The Halden Task Complexity experiments looked at different factors that would increase the probability of human failures such as the effects of time pressure/information load and masked events. EPRI and ATHEANA could use the design of the Halden experiments as a model for future simulations because they produced results that showed important differences in crew performance under certain conditions. Both models can also use the Halden experiments and results to sensitize the experts and analysts to the real effects of an error forcing context. / by Mary R. Presley. / S.M.and S.B.

Thesis: Sc. D., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 2014. / This electronic version was submitted by the student author. The certified thesis is available in the Institute Archives and Special Collections. / Cataloged from student-submitted PDF version of thesis. / Includes bibliographical references. / High-performance operation in tokamaks is characterized by the formation of a pedestal, a region of suppressed transport and steep gradients in density, temperature, and pressure near the plasma edge. The pedestal height is strongly correlated with overall fusion performance, as a substantial pedestal supports the elevated core pressure necessary for the desired fusion reaction rate and power density. However, stationary operation requires some relaxation of the particle transport barrier, to avoid the accumulation of impurities (e. g., helium "fusion ash," plasmafacing surface materials) in the plasma. Moreover, the formation of the pedestal introduces an additional constraint: the steep gradients act as a source of free energy for Edge-Localized Mode (ELM) instabilities, which on ITER- or reactor-scale devices can drive large, explosive bursts of particle and energy transport leading to unacceptable levels of heat loading and erosion damage to plasma-facing materials. As such, the suppression, mitigation, or avoidance of large ELMs is the subject of much current research. In light of this, a firm physical understanding of the pedestal structure and stability against the ELM trigger is essential for the extrapolation of high-performance regimes to large-scale operation, particularly in operating scenarios lacking large, deleterious ELMs. This thesis focuses on the I-mode, a novel high-performance regime pioneered on the Alcator C-Mod tokamak. I-mode is unique among high-performance regimes in that it appears to decouple energy and particle transport, reaching H-mode levels of energy confinement with the accompanying temperature pedestal while maintaining a L-mode-like density profile and particle transport. I-mode exhibits three attractive properties for a reactor regime: (1) I-mode appears to be inherently free of large ELMs, avoiding the need for externally-applied ELM control. (2) The lack of a particle transport barrier maintains the desired level of impurity flushing from the plasma, avoiding excessive radiative losses. (3) Energy confinement in I-mode presents minimal degradation with input heating power, contrary to that found in H-mode. This thesis presents the results from a combined empirical and computational study of the pedestal on C-Mod. Analysis methods are first implemented in ELMy H-mode base cases on CMod -- in particular, the EPED model based on the combined constraints from peeling-ballooning MHD instability and kinetic-ballooning turbulence is tested on C-Mod. Empirical results in ELMy H-mode are consistent with the physics assumptions used in EPED, with the pedestal pressure gradient constrained by [delta]p ~ I2/p expected from the ballooning stability limit. To lowestorder approximation, ELMy H-mode pedestals are limited in [beta]p,ped, with the attainable beta set by shaping -- within this limit, an inverse relationship between pedestal density and temperature is seen. The pedestal width is found to be described by the scaling [delta][psi] = G[beta] 1/2 / p.ped expected from the KBM limit, where G([nu],[epsilon], ...) is a weakly varying function with hGi = 0.0857. No systematic secondary scalings with field, gyroradius, shaping, or collisionality are observed. The EPED model, based on these assumptions, correctly predicts the pressure pedestal width and height to within a systematic ~20% uncertainty. Empirical scalings in I-mode highlight the operational differences from conventional H-modes. The temperature and pressure pedestal exhibit a positive trend with current, similar to H-mode (although I-mode pedestal temperature typically exceeds that found in comparable H-modes) -- however, the temperature and pressure respond significantly more strongly to heating power, with Te ... The I-mode density profile is set largely independently of the temperature pedestal (unlike ELMy H-mode), controlled by operator fueling. Given sufficient heating power to maintain a consistent ..., temperature pedestals are matched across a range of fueling levels. This indicates a path to readier access and increased performance in Imode, with the mode accessed at moderate density and power, after which the pedestal pressure is elevated with matched increases in fueling and heating power. Global performance metrics in I-mode are competitive with H-mode results on C-Mod, and are consistent with the weak degradation of energy confinement with heating power. I-mode pedestals are also examined against the physics basis for the EPED model. Peelingballooning MHD stability is calculated using the ELITE code, finding the I-mode pedestal to be strongly stable to the MHD modes associated with the ELM trigger. Similarly, modeling of the KBM using the infinite-n ballooning mode calculated in BALOO as a surrogate for the threshold indicates that the I-mode pedestal is stable to kinetic-ballooning turbulence, consistent with the observed lack of a trend in the pedestal width with [beta]p,ped. This is found to be the case even in I-modes exhibiting small, transient ELM-like events. The majority of these events are triggered by the sawtooth heat pulse reaching the edge, and do not negatively perturb the temperature pedestal -- it is proposed, then, that these events are not true peeling-ballooning-driven ELMs, but rather are an ionization front in the SOL driven by the sawtooth heat pulse. There are transient ELM events showing the characteristic temperature pedestal crash indicating a true ELM -- the steady I-mode pedestals around these isolated events are also modeled to be P-B and KBM stable, although more detailed modeling of these events is ongoing. / by John Reel Walk, Jr. / Sc. D.

Thesis (S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2012. / Cataloged from PDF version of thesis. / Includes bibliographical references (p. 140-148). / The main drawback to renewable energy systems is the higher cost of production compared to competitors such as fossil fuels. Thus, there is a need to increase the efficiency of renewable energy systems in an effort to make them more cost competitive. In this study, the use of nanosurfaces is evaluated for its benefits in improving the efficiency of a concentrated solar tower power system by increasing the energy retained by the receiver surface, and for reducing the fouling on geothermal heat exchangers. The samples tested for the solar receiver application were Inconel 617, Inconel 617 with a 150 nm layer of platinum, Inconel 617 with a 150 nm layer of platinum and a 550 nm layer of nickel oxide, oxidized nickel, and silicon carbide. The experimental results indicated that the platinum was an ineffective diffusion barrier, nickel oxide displays solar selective properties, and silicon carbide would be the best choice for a surface among the samples tested. This indicates that at the operating temperatures for this receiver at 700 °C, a black body surface is more effective than a practical solar selective surface. The nanosurfaces tested for the antifouling application in geothermal systems were subjected to chemistry conditions similar to that in a Dry Cooling Tower at a geothermal plant in Larderello, Italy. Each sample's performance was measured by determining each samples weight change and surface characterization after exposure in an experimental loop. The best performing coatings, all of which showed negligible weight gain, were the Curran 1000 coating from Curran International, the Curran 1000 coating with nanographene, and the Curralon coating with PTFE. Upon further analysis, the Curran 1000 with nanographene was identified as the most promising coating option. / by Alexander W. Rehn. / S.M.

Thesis: S.M., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 2017. / This electronic version was submitted by the student author. The certified thesis is available in the Institute Archives and Special Collections. / Cataloged from student-submitted PDF version of thesis. / Includes bibliographical references (pages 119-125). / The nitrogen-vacancy (NV) center in diamond has emerged as a promising platform for high-sensitivity, vector magnetic field detection and high spatial resolution magnetic-field imaging due to its unique combination of optical and spin properties. NV diamond magnetometry has enabled a wide array of applications from the noninvasive measurement of a single neuron action potential to the mapping [mu]T-fields in [mu]m-size meteorite grains. To further improve the magnetic sensitivity of an ensemble NV magnetometer, the growth and processing of the host diamond must be taken into account. This thesis presents a systematic study of the effects of diamond processing on bulk chemical-vapor-deposition diamond. In particular, NV charge-state composition and spin decoherence times are measured for diamonds irradiated with 1 MeV electrons at doses of 1x1015-5x1019 e-/cm2 and thermally annealed at temperatures of 850°C and 1250°C. The study provides an optimal range for diamond processing and shows the quenching of the NV center at high irradiation dosage from the creation of additional vacancy-related defects. / by Scott T. Alsid. / S.M.