Thesis (Ph. D.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2008. / Cataloged from PDF version of thesis. / Includes bibliographical references (p. 243-250). / Large superconducting magnets used in fusion reactors, as well as other applications, need a diagnostic that can non-invasively measure the temperature and strain throughout the magnet in real-time. A new fiber optic sensor has been developed for these long-length superconducting magnets that simultaneously measures the temperature and strain based on spontaneous Brillouin scattering in an optical fiber. Using an extremely narrow (200 Hz) linewidth Brillouin laser with very low noise as a frequency shifted local oscillator, the frequency shift of spontaneous Brillouin scattered light was measured using heterodyne detection. A pulsed laser was used to probe the fiber using Optical Time Domain Reflectometry (OTDR) to define the spatial resolution. The spontaneous Brillouin frequency shift and linewidth as a function of temperature agree well with previous literature of stimulated Brillouin data from room temperature down to 4 K. Analyzing the frequency spectrum of the scattered light after an FFT gives the Brillouin frequency shift, linewidth. and intensity. For the first time, these parameters as a function of strain have been calibrated down to 4 K. Measuring these three parameters allow for simultaneously determining the temperature and strain in real-time throughout a fiber with a spatial resolution on the order of several meters. The accuracy of the temperature and strain measurements vary over temperature-strain space, but an accuracy of better than + 2 K and ± 100 Pe are possible throughout most of the calibrated temperature-strain space (4-298 K and 0-3500 p/g). In the area of interest for low-temperature superconducting magnets (4-25 K), the temperature accuracy is better than + 1 degree. / (cont.) This temperature accuracy, along with the sub-second measurement time, allows this system to be used not only as a quench detection system, but also as a quench propagation diagnostic. The sensing fiber can also simultaneously provide the first ever spatially resolved strain measurement in an operating magnet. / by Scott Brian Mahar. / Ph.D.

Thesis (Ph. D.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2011. / Cataloged from PDF version of thesis. "June 2011." / Includes bibliographical references. / In response to increasing demands for the services of research reactors, a 5 MW LEU-fueled research reactor core is developed and optimized to provide high thermal flux within specified limits upon thermal hydraulic performance, cycle length, irradiation utilization, and manufacturability. A novel fuel assembly concept which makes use of integral flux traps is postulated to meet these requirements. Each assembly can be rotated into one of three different configurations to produce flux traps of different size, shape, and neutron energy spectrum within the core. A method for predicting and guiding the search for the optimum geometry was sought. Kriging has been chosen to predict the values of eigenvalue and thermal flux at untested geometric parameters. Because kriging treats all measurements as the sum of a global deterministic function and a stochastic departure from that function, predictions come with a measurement of uncertainty. As a result, the analyst can search the design space for likely improvement, or probe areas of high uncertainty for improvements that might have been missed using other methods. The technique is used in an algorithm for constrained optimization of the design, and a set of best practices for use of this are described. The optimized design produces a peak thermal flux of 1.56 x 10¹⁴ n/cm²s. Safety is demonstrated by presentation of reactivity feedback coefficients and the results of loss of flow and reactivity insertion transient analysis. A single fission target can be used to produce 96 6-day Ci of ⁹⁹Mo per week. When the reactor is oriented to take advantage of high fast flux, steels can be subjected to damage rates of 5.76 dpa per year. Silicon carbide can be damaged at a rate of 2.79 dpa/y. The concept is a safe, versatile, proliferation-resistant means of supplying current and future irradiation needs. / by Stephanie Anne Kempf. / Ph.D.

Thesis (Ph. D.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2013. / Cataloged from PDF version of thesis. Page 224 blank. / Includes bibliographical references. / Fast ions generated from laser-plasma interactions (LPI) have been used to study inertial confinement fusion (ICF) implosions and laser-foil interactions. LPI, which vary in nature depending on the wavelength and intensity of the driver, generate hot electrons with temperatures ranging from tens to thousands of kilo-electron-volts. These electrons, which accelerate the ions measured in this work, can be either detrimental or essential to implosion performance depending on the ICF scheme employed. In direct-drive hot-spot ignition, hot electrons can preheat the fuel and raise the adiabat, potentially degrading compression in the implosion. The amount of preheat depends on the hot-electron source characteristics and the time duration over which electrons can deposit energy into the fuel. This time duration is prescribed by the evolution of a sheath that surrounds the implosion and traps electrons. Fast-ion measurements have been used to develop a circuit model that describes the time decay of the sheath voltage for typical OMEGA implosions. In the context of electron fast ignition, the produced fast ions are considered a loss channel that has been characterized for the first time. These ions have also been used as a diagnostic tool to infer the temperature of the hot electrons in fast-ignition experiments. It has also been shown that the hot-electron temperature scales with laser intensity as expected, but is enhanced by a factor of 2-3. This enhancement is possibly due to relativistic effects and leads to poor implosion performance. Finally, fast-ion generation by ultra-intense lasers has also been studied using planar targets. The mean and maximum energies of protons and heavy ions has been measured, and it has been shown that a two-temperature hot-electron distribution affects the energies of heavy ions and protons. This work is important for advanced fusion concepts that utilize ion beams and also has applications in medicine. / by Nareg Sinenian. / Ph.D.

Thesis: Ph. D., 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 176-181). / In the fuel assemblies of a boiling water reactor (BWR) the steam quality increases along the assembly's length as heat is transferred from the fuel rods to the water coolant. Nucleate boiling is the dominant heat transfer mechanism at low and intermediate steam qualities (typical of the bubbly and slug/churn flow regimes), while forced convective evaporation dominates at higher steam quality in the annular flow regime. The transition of the heat transfer mechanism, also called suppression of nucleate boiling (SNB), affects the local heat transfer coefficient (HTC), the stability of the liquid film, and the entrainment dynamics. To support the efficient design and safe operation of future BWRs with higher power density, a thorough understanding of the thermohydraulic mechanisms and an accurate prediction of the transition conditions for SNB in annular flow is quite desirable. An innovative diagnostic technique combining synchronized infrared thermography and an electrical conductance-based liquid film thickness sensor was utilized here to investigate the details of the SNB phenomena with high spatial and temporal resolutions. The main control parameters of the tests included: the mass flux from 700 to 1400 kg-m⁻²-s⁻¹, steam quality from 0.01 to 0.08, and heat flux from 100 to 2000 kW-m⁻². The system pressure was held close to atmospheric pressure. At each set of conditions, the local distributions of the 2D surface temperature, 2D heat flux, and quasi-2D liquid film thickness were measured. From the measured data, the SNB heat flux, the SNB wall superheat, and the hydrodynamic properties of the disturbance waves were extracted. The experimental observations show for the first time the multiscale interaction of the extremely thin film and small nucleation cavities (on the scale of 10 micron), with the large disturbance waves and their associated temperature oscillations (with wavelengths of ~10 cm). A first of a kind 1D mechanistic model was developed to accurately capture this unique transient effect of the disturbance waves on the local heat transfer. The experimental results also suggest a strong dependency of the SNB heat flux and wall superheat on steam quality, with a second-order, weaker dependency on total mass flux. The same dependency is also found for the disturbance wave properties. A complete set semi-empirical correlations was proposed for predicting the time-averaged film thickness and SNB thermal conditions. Good agreement is found between the semi-empirical correlations and the experimental results. The database generated in this project can be further used for development and validation of CFD models of SNB and two-phase heat transfer in annular flow. / by Guanyu Su. / Ph. D.

Thesis (Ph. D.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, February 2006. / Includes bibliographical references (p. 229-232). / One of the ways to lower the cost of nuclear energy is to increase the power density of the reactor core. Features of fuel design that enhance the potential for high power density are derived based on characteristics of the pressurized water reactor (PWR) and its related design limits. Those features include: large fuel surface to volume ratio, small fuel thickness, large fuel rod stiffness, low core pressure drop and an open fuel lattice design. Three types of fuel designs are evaluated from the thermal-hydraulic point of view: conventional solid cylindrical fuel rods, internally and externally cooled annular fuel rods, and spiral cross-geometry fuel rods, with the major effort allocated to analyzing the annular fuel. Limits of acceptable power density in solid cylindrical fuel rods are obtained by examining the effects of changing the core operation parameters, fuel rod diameter and rod array size. It is shown that the solid cylindrical geometry does not meet all the desired features for high power density well, and its potential for achieving high power density is limited to 20% of current PWR power density, unless the vibration problems at the coolant higher velocity are overcome. The internally and externally cooled annular fuel potential for achieving high power density is explored, using a whole core model. The best size of fuel rods that fits in the reference assembly dimension is a 13x13 array, since the hot red will have a balanced MDNBR in the inner and outer channels. With proportional increase in coolant flow rate, this annular fuel can increase PWR power density by 50% with the same DNBR margin, while reducing by 1000 'C the peak fuel temperature. Five issues involving manufacturing tolerances, oxide growth on rod surfaces, inner and outer gap conductances asymmetry, MDNBR sensitivity to changes in core operation parameter and resistance to instabilities were also evaluated. / (contd.) It is found that the main uncertainty for this design is associated with the heat split between the inner and outer channels due to differences in the thermal resistances in the two fuel-clad gaps. Annular fuel is found to be resistant to flow instabilities, such as Ledinegg instability and density wave oscillation due to high system pressure and one-phase flow along most of the hot channel length. Similar power density uprate is found possible for annular fuel in a hexagonal lattice. Large break loss of coolant accident (LBLOCA) for the reference Westinghouse 4-loop PWR utilizing annular fuel at 150% power is analyzed using RELAP, under conservative conditions. The blowdown peak cladding temperature (PCT) is found to be lower because of the low operating fuel temperature, but the flow rate from the safety injection system needs to be increased by 50% to remove the 50% higher decay heat. Loss of flow analysis also showed better performance of the annular fuel because of its low stored energy. The fuel design that best meets the desired thermal and mechanical features is the spiral cross-geometry rods. The dimensions of this type of fuel that can be applied in the reference core were defined. Thermal-hydraulic whole-core evaluations were conducted with cylindrical fuel rod simplification, and critical heat flux modification based on the heat flux lateral non-uniformity in the cross geometry. This geometry was found to have the potential to increase PWR power density by 50%. However, there are major uncertainties in the feasibility and costs of manufacturing this fuel. / by Dandong Feng. / Ph.D.

Thesis (Ph. D. in Applied Plasma Physics)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2012. / Cataloged from PDF version of thesis. / Includes bibliographical references (p. 213-220). / In this thesis we discuss measurements and modeling of minority heated fast-ion distributions in the Ion Cyclotron Range of Frequencies (ICRF) on the Alcator C-Mod tokamak. Analysis of fast-ions >100Te is important for both ITER and a future fusion reactor as both will have a significant population of 3.5 MeV alpha particles generated in fusion reactions. Fast particles in this energy range can drive unstable modes such as Toroidal Alfvén Eigenmodes (TAEs) and Reversed Shear Alfvén Eigenmodes (RSAEs). Furthermore, energetic ions may display plasma properties that differ from the bulk plasma. It is crucial to benchmark current simulation codes with measurements from highly energetic fast-ions on current devices. This thesis will focus on measurements of the fast-ion distribution made on C-Mod with an upgraded Compact Neutral Particle Analyzer (CNPA). Measurements of the fast-ion distributions will reveal strong dependences of the fast-ion effective temperature on both electron density and plasma current. For further analysis, we use the simulated distributions generated by the coupled full-wave spectral solver AORSA, with the zero orbit-width bounce-averaged Fokker-Planck code CQL3D. A new synthetic diagnostic integrated into CQL3D is used to make direct comparisons with the CNPA. We find that for plasmas that have a steady-state fast-ion distribution (df /dt = 0) the simulation and the experiment have good agreement. However, in simulations where the fast-ion distribution is evolving in time (df/dt =/ 0) we find a discrepancy between the simulation and the experimental results. The simulation is seen to evolve much slower than the experiment. Various reasons for the discrepancy are explored, including the possibility of a violation of the quasi-linear theory used in CQL3D. / by Aaron Craig Bader. / Ph.D.in Applied Plasma Physics

Thesis: S.M., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 2014. / Cataloged from PDF version of thesis. / Includes bibliographical references (pages 83-84). / Grid-to-rod fretting (GTRF) has been the major cause of fuel leakage in Pressurized Water Reactors (PWRs) for the past ten years. It is responsible for over 70% of the fuel leaking in PWRs in the United States. The Consortium for Advanced Simulation of Light Water Reactors (CASL) has identified GTRF as one of the "Challenge Problems" that motivates the need for development and application of a modeling environment for predictive simulation of light water reactors. In this thesis, an initial verification of the Fluid Structure interaction (FSI) coupling algorithm for flow inside a vibrating tube was conducted using CFD software STAR CCM+. The benchmark confirmed accurate predictions of the coupled frequencies over a wide range of Reynolds numbers, providing good confidence on the generality of the approach. A representative spacer model was then developed to be used to evaluate the coupling phenomena in GTRF applications. The geometry consists of a 2 span, 3x3 spacer grid. To create the coupled fluid-solid test geometry, a solid Zircaloy cladding was added to the geometry. The solid cladding was added to capture fluid structure interaction effects. The spacer grid supports were altered to mimic having experienced relaxation and allowing free movement of the fuel rod for small displacements. A desirable mesh was constructed over the geometry. Large Eddy Simulations (LES) have been performed to accurately compute the turbulent forces acting on the spacers. Simulations were first performed for a rigid rod, as a reference decoupled solution. Fully coupled simulations were successively performed allowing for the evaluation of the complexity of the fluid-structure coupling behavior. Results of the simulations were also compared to previous Westinghouse analysis performed on a production spacer with a decoupled approach, to confirm the prototypical performance of the geometrical configuration adopted in the present work. The ultimate goal of this thesis was to demonstrate the practicability of a fully coupled FSI simulation for PWR fuel simulations, and further to advance the understanding of the complex fluid structure coupling in PWR fuel assemblies. / by Sasha Angela Tan-Torres. / S.M.

Thesis: Ph. D., 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 389-400). / Zircaloy-4 has been used in pressurized water reactors (PWRs) for decades, and enhanced corrosion rates in reactors compared to out-of-pile have long been observed. However, the exact mechanism explaining the early departure from autoclave kinetics after 3-5 microns of oxide have formed has proved elusive. This thesis considers and evaluates a number of possible explanations for this early acceleration in kinetics. The bulk of the evidence points to Fe depletion from secondary phase particles (SPPs) as the culprit in enhancing Zircaloy-4 corrosion rates in PWRs. These new findings have been incorporated in a mechanistic finite-element phase field model of Zircaloy-4 corrosion called HOGNOSE. It accounts for both diffusion-and drift-based oxygen anion transport in Zircaloy-4 by including the effects of radiation-induced evolution of SPPs in changing the contribution of a local charge transport inequality through their depletion and release of iron. By addressing the imbalance in charged particle transport, the code can be adapted to model multiple zirconium-based alloys in autoclave and irradiated conditions with minimal parameter fitting. Rather than the typical empirical approach, HOGNOSE uses a physics-based methodology to capture the early agreement between autoclave and in-reactor data and the point at which reactor kinetics are enhanced compared to autoclave kinetics. HOGNOSE results agree fairly well with those observed in experiments for oxide thicknesses less than 10 microns, above which other enhancement mechanisms can no longer be safely ignored. HOGNOSE captures increasing amorphization with decreasing temperature, and more subtle corrosion rate enhancement at higher temperatures. Comparisons between HOGNOSE results and literature data suggest that the next focus for mechanistic modeling should consider additional neutron flux effects. To support HOGNOSE development, corrosion testing of Zircaloy-4 in steam at atmospheric pressure and 415 degrees Celsius was performed. Samples were analyzed using a focused ion beam/scanning electron microscope (FIB/SEM) to obtain oxide thickness measurements with greater temporal resolution than is widely provided by autoclave testing. Oxide thickness data was used to determine the thermal dependence of oxygen diffusivity in the oxide within HOGNOSE. HOGNOSE would also benefit from measurements of the concentrations and charge states of cation dopants in post-irradiated Zircaloy oxides to help determine whether this model is truly accurate in its physical description. / by Andrew Frederic Dykhuis. / Ph. D.

Thesis (S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2008. / Includes bibliographical references (p. 53-55). / Combating the threat of nuclear smuggling through shipping ports and border crossings has been recognized as a national priority in defending the US against nuclear terrorism. In light of the SAFE Port act of 2006, the Domestic Nuclear Detection Office (DNDO) has been charged with the responsibility of providing the Customs and Border Protection Agency (CBP) with the capability to conduct 100% radiological screening of all containers entering the country. In an attempt to meet this mandate, the DNDO has conducted a typical government acquisition procedure to develop and acquire radiation portal monitors (RPMs) capable of passive gamma-ray spectroscopy that would allow 100% radiological screening without detrimental affects on the stream of commerce through the terminals. However, the Cost-Benefit Analysis (CBA) supporting the DNDO decision-making process has been criticised and has delayed the program significantly. We propose an Analytic-Deliberative Process (ADP) as an alternative to CBA for this application. We conduct a case study with four DNDO stakeholders using the ADP proposed by the National Research Council in the context of environmental remediation and adapted by the MIT group and compare the results to those derived from DNDO's CBA. The process involves value modeling using an objectives hierarchy and the analytic hierarchy process. Value functions are derived and expected outcomes for the decision options are elicited from the stakeholders. The process results in a preference ranking of the decision options in order of value to each stakeholder. The analytical results are then used to structure a deliberation in which the four stakeholders use both the analytical results and any pertinent information outside the analysis to form a consensus. / (cont.) The final decision of both the CBA and ADP models show good agreement and demonstrate the validity of both methods. However, the ADP format is better at explicitly capturing and quantifying subjective influences affecting the final decision. This facilitates discussion and leads to faster consensus building. / by Michael B. Shattan. / S.M.

Thesis (S.M. and S.B.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2009. / Cataloged from PDF version of thesis. / Includes bibliographical references (p. 92). / Many desirable locations for Pebble Bed Modular Reactors are currently out of consideration as construction sites for current designs due to limitations on the mode of transportation for large RPVs. In particular, the PBMR-400 design developed by PBMR Pty of South Africa uses an RPV with an outer diameter of 6.4 meters. Since current SCHNABEL railcars can only haul components up to 4.3 meters wide, the only other possibility for transport is by barge, which limits construction to sites accessible by river, lake or coast. Designing a PBMR with a core able to fit within an RPV able to be transported by railcar would be extremely valuable, especially for potential inland sites only accessible by railway, such as those in the Canadian Oil Sands at which the PBMR would be utilized for oil extraction processes. Therefore, a study was conducted to determine the feasibility of a Pebble Bed Modular Reactor design operating at 250 MWth with a core restricted to fitting inside an RPV with an outer diameter of 4.3 meters. After reviewing the performance of various core configurations satisfying this constraint, an optimized PBMR design operating at this power was found. This new design uses the same fuel management scheme as the PBMR 400, as well as similar inlet and outlet coolant temperatures. This MPBR-250 design includes a pebble bed with an outer diameter of 2.7 meters, an outer reflector 50 cm thick and 12.5% enriched fuel. A mixture of graphite pebbles of 11.7% is also included in the pebble bed to produce an equilibrium core with minimal excess reactivity. / (cont.) This thesis shows that the MPBR-250 can perform up to the standards of the PBMR-400 design with respect to power peaking factors, peak temperatures and RPV fast fluences and can also increase fuel burnup to nearly 110 GWd/T. In addition, the MPBR-250 is a much more agile design, able to be deployed at a wider variety of locations because its RPV can be transported by railcar. / by Matthew S. Everson. / S.M.and S.B.