Thesis: S.M., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 2016. / Cataloged from PDF version of thesis. / Includes bibliographical references (pages 125-132). / SILEX is a molecular isotope separation technique that takes advantage of the differing energies of molecular excitations between different isotopes of uranium. This process occurs within a jet of supersonic gas, the gas includes both uranium hexafluoride and a carrier gas, and reportedly allows for high separation factors relative to other isotope-separation processes. Industry interests have argued that it could be readily commercialized. This topic is of particular interest because laser isotope separation technology has seen an increase in interest and funding over the last decade. This suggests some study of the risks that such a technology poses to society may now be in order. To inform policymakers about the risks inherent to a particular enrichment technology, it is necessary to understand the theoretical underpinnings of the technology before one can analyze the impact of the technology. Positions expressed in the current literature are ill-informed and range from deep opposition, citing concerns that SILEX poses greater proliferation risk than centrifuge or gaseous diffusion technology, to claims by scientists that it is not possible to use the technology to produce greater than 50% enriched U-235. A rigorous and holistic view of the technology will better inform policy by improving the accuracy of claims and identifying realistic solutions to problems the technology may pose. This thesis will seek to provide this deeply technical and holistic analysis of the technology, and will use the results to interpret the economic and proliferation impact such a technology will have on the global nuclear enterprise. The holistic analysis in this thesis will present several important conclusions: 1) the enrichment factor of SILEX is not quite as large as proponents suggest; 2) asymmetric cascade designs will be required; 3) SILEX may not be cheaper than centrifuge facilities; 4) SILEX will not be viable without improvements in laser technology; and 5) international policies may be the most effective means of curtailing enrichment schemes like SILEX. / by Aaron Taylor Baldwin. / S.M.

Thesis (S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2013. / Cataloged from PDF version of thesis. / Includes bibliographical references (pages 139-147). / The two-color interferometer diagnostic on Alcator C-Mod has been upgraded to measure line-integrated electron density fluctuations for turbulence and transport studies. Heterodyne signals from ten vertical-viewing CO2 laser chords are demodulated relative to a local oscillator using high bandwidth analog in-phase/quadrature electronics, replacing lower bandwidth digital fringe counting electronics. The raw outputs of the high bandwidth electronics, which are proportional to the sine and cosine of the interferometric phase shift, are digitized at up to 10 MHz, which is sufficient for fluctuation analysis. Extraction of the measured phase from the sine and cosine signals is now performed entirely in software, providing the line-integrated electron density at high bandwidth. The interferometer design, calibration, and measurement uncertainty is presented. Measurement uncertainties due to nonlinearities in the analog electronics are found to be comparable to the uncertainties of the previous system. The interferometer can now resolve line-integrated electron density fluctuations with major-radial wavenumbers below ... . The new fluctuation measurement capability is used to partially verify the calibration and low-kR wavenumber response of phase-contrast imaging, to aid in gyrokinetic transport model validation research. Agreement between the two diagnostics is demonstrated for broadband fluctuations and the low-kR component of the quasi-coherent mode, improving confidence in the calibration of the phase-contrast imaging system. Both diagnostics observe a series of fluctuations during quasi-steady periods of minority heated I-mode plasmas with strong off-axis electron heating. The observed fluctuations are localized to the plasma core using Doppler shift analysis and data from edge fluctuation diagnostics. Transport analysis shows that the fluctuations do not correlate with enhanced thermal transport, and gyrokinetic linear stability analysis shows that the plasma is stable to drift wave turbulence, ruling out the possibility that the observed fluctuations are destabilized drift wave turbulence. / by Cale Phillip Kasten. / S.M.

Thesis: S.B., 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 36-38). / Application and fundamental understanding of magnetic materials requires knowledge of their properties and phase diagrams, which depend on the structure and dynamics of magnetic domains. Current techniques for studying magnetic domains are limited to imaging near-surface magnetic structure. Techniques for investigating bulk magnetic structure are under development, particularly for applications in electronics and superconductivity. This study investigates the feasibility of observing bulk magnetic domain structure using polarized neutron imaging and tomography. Polarized neutrons are advantageous for studying internal magnetic structure because they penetrate materials, and their spin-polarizations are sensitive to magnetic fields. This study experimentally tested the depolarization of neutrons in five different materials with known magnetic order using PONTO, an instrument at Helmholtz-Zentrum Berlin (Germany) that uses a polarization-sensitive filter and analyzer to measure neutron depolarization in magnetic samples. Magnetic structure was observed when samples were subjected to magnetic fields to increase magnetization, and to cooling beneath the Curie temperature. Samples measured at zero field and room temperature randomly depolarized the neutron beam because their domains are smaller than the resolution of PONTO. Successful observation of magnetic effects indicates the promise of polarized neutron imaging for studying bulk magnetic domain structure; however, further development of imaging methods is necessary for understanding the connection between neutron depolarization and domain structure. / Alexandra R. Delmore. / S.B.

Thesis (Ph. D.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2012. / Cataloged from PDF version of thesis. / Includes bibliographical references (p. 203-208). / In this thesis, we investigate the potential role played by large-scale anomalous heat sources (e.g. prolonged heat wave events) in generating acoustic-gravity waves (AGWs) that might trigger widespread plasma turbulence in the ionospheric layer. The main hypothesis is that, the thermal gradients associated with the heat wave fronts could act as a source of powerful AGW capable of triggering ionospheric plasma turbulence over extensive areas. In our investigations, first we are going to examine a case study of the summer 2006 North American heat wave event. Our examination of GPS-derived total electron content (TEC) data over the North American sector reveals a quite noticeable increase in the level of daily plasma density fluctuations during the summer 2006 heat wave period. Comparison with the summer 2005 and summer 2007 data further confirms that the observed increase of traveling ionospheric disturbances (TIDs) during the summer 2006 heat wave period was not simply a regular seasonal phenomenon. Furthermore, a series of field experiments had been carried out at the High-frequency Active Auroral Research Program (HAARP) facility in order to physically simulate the process of AGW/TID generation by large-scale thermal gradients in the ionosphere. In these ionospheric HF heating experiments, we create some time-varying artificial thermal gradients at an altitude of 200-300 km above the Earth's surface using vertically-transmitted amplitude-modulated 0-mode HF heater waves. For our experiments, a number of radio diagnostic instruments had been utilized to detect the characteristic signatures of heater-generated AGW/TID. So far, we have been able to obtain several affirmative indications that some artificial AGW/TID are indeed being radiated out from the heated plasma volume during the HAARP-AGW experiments. Based on the experimental evidence, we may conclude that it is certainly quite plausible for large-scale thermal gradients associated with severe heat wave events to generate some AGW which might induce widespread plasma turbulence far in space. / by Rezy Pradipta. / Ph.D.

Thesis: S.M., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 2015. / 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 (page 88). / An EPRI study[1], published in 2011, used measured flux map data (taken over 44 operational cycles of the Catawba and McGuire nuclear power plants) to determine fuel assembly reactivity decrements versus burnup. The analytical techniques used to infer measured assembly reactivities required perturbation calculations using 3D nodal diffusion core models. Subsequently, questions have arisen within the Nuclear Regulatory Commission (NRC) as to potential uncertainties in measured assembly reactivity decrements that might have arisen from approximations of the 2-group nodal methods and perturbation techniques employed. Subsequently, Gunow[2] used full-core, multi-group, neutron transport models to replace the nodal diffusion core models, and he demonstrated that measured reactivity decrements were independent of the core model. In this thesis, two cycles of the BEAVRS PWR reactor benchmark are used to test the EPRI methodology, now including testing of not only the nodal core diffusion model, but also the perturbation technique itself. By changing the perturbation technique from assembly reactivity to assembly-average fuel temperature, it is demonstrated that measured reactivity decrements are almost independent of the perturbation technique - with a level of precision greater then the 250 pcm reactivity decrement uncertainty assigned in the EPRI study. These new results demonstrate that the reactivity decrements and uncertainties derived by nodal diffusion and burnup perturbation in the original EPRI study hold up to further scrutiny, and they remain credible for licensing application of burnup credit in Spent Fuel Pool (SFP) criticality analysis. / by Elliot M. Sykora. / S.M.

Thesis (S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2007. / Includes bibliographical references (p. 199-201). / Preliminary analysis of the cross-shaped spiral (CSS) fuel assembly suggests great thermal-hydraulic upside. According to computational models, the increase in rod surface area, combined with an increase in coolant turbulence and inter-channel mixing will allow for a greater than 25% uprate in total core power, without loss of safety margin. Proper design of the rod dimensions can limit circumferential heat-flux to a peak-to-average ratio of 1.88. Non-uniformities in heat flux due to its unusual geometry seem to particularly ally CSS fuel to the BWR core, where limiting conditions are less likely to be locally influenced. Furthermore, the increase in cooling surface and reduction in central pin thickness is expected to drop fuel centerline temperature an estimated 2000C under nominal operating conditions, a reduction which rises to 3000C at 125% of nominal power conditions. In addition to these advantages, the absence of grid spacers within the CSS fuel assembly is expected to lower pressure losses, aiding natural convection and core stability. Spacers typically account for 25-30% of the total core pressure drop. Experimental measurements of hydraulic: losses for 1.5-meter-long model CSS rods in 4x4 arrays show a larger pressure drop at the same flow velocity than for bare cylindrical rods. However, this results in a CSS-bundle turbulent friction factor which is only 90% of the expected value given its hydraulic diameter. The effect of twist pitch on this pressure drop and friction factor is negligible in the range of twists examined. / (cont.) Combined with the elimination of grid spacers, this results in a 40% reduction in core hydraulic loss from the reference case (neglecting entrance and exit plates). All told, the use of CSS rods should reduce total core pressure drop at nominal power by 9%, in spite of a reduction in core flow area. At 125% of nominal power, this becomes a 16% increase in pressure drop in comparison to the reference core at nominal power. / by Thomas M. Conboy. / S.M.

Thesis (Ph. D.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2010. / Cataloged from PDF version of thesis. / Includes bibliographical references (p. 278-285). / Density Wave Oscillations (DWOs) are known to be possible when a coolant undergoes considerable density reduction while passing through a heated channel. In the development of boiling water reactors (BWRs), there has been considerable concern about the effects of such oscillations when coupled with neutronic feedback. The current trend of increasing reactor power density and relying more extensively on natural circulation for core cooling may have consequences for the stability characteristics of new BWR designs. This work addresses a wide range of issues associated with the BWR stability: 1) flashing-induced instability and natural circulation BWR startup; 2) stability of the BWRs with advanced designs involving high power :densities; 3) modeling assumptions in stability analysis methods; and 4) the fuel clad performance during power and flow oscillations. To capture the effect of flashing on density wave oscillations during low pressure startup conditions, a code named FISTAB has been developed in the frequency domain. The code is based on a single channel thermal-hydraulic model of the balance of the water/steam circulation loop, and incorporates the pressure dependent water/steam thermodynamic properties, from which the evaporation due to flashing is captured. The functionality of the FISTAB code is confirmed by testing the experimental results at SIRIUS-N facility. Both stationary and perturbation results agree well with the experimental results. The proposed ESBWR start-up procedure under natural convection conditions has been examined by the FISTAB code. It is confirmed that the examined operating points along the ESBWR start-up trajectory from TRACG simulation will be stable. To avoid the instability resulting from the transition from single-phase natural circulation to two-phase circulation, a simple criterion is proposed for the natural convection BWR start-up when the steam dome pressure is still low. Using the frequency domain code STAB developed at MIT, stability analyses of some proposed advanced BWRs have been conducted, including the high power density BWR core designs using the Large Assembly with Small Pins (LASP) or Cross Shape Twisted (CST) fuel designs developed at MIT, and the Hitachi's RBWR cores utilizing a hard neutron spectrum and even higher power density cores. The STAB code is the predecessor of the FISTAB code, and thermodynamic properties of the coolant are only dependent on system pressure in STAB. It is concluded that good stability performance of the LASP core and the CST core can be maintained at nominal conditions, even though they have 20% higher reactor thermal power than the reference core. Power uprate does not seem to have significant effects on thermal-hydraulic stability performance when the power-to-flow ratio is maintained. Also, both the RBWR-AC and RBWR-TB2 designs are found viable from a stability performance point of view, even though the core exit qualities are almost 3 times those of a traditional BWR. The stability of the RBWRs is enhanced through the fast transient response of the shorter core, more flat power and power-to-flow ratio distributions, less negative void feedback coefficient, and the core inlet orifice design. To examine the capability of coupled 3D thermal-hydraulics and neutronics codes for stability analysis, USNRC's latest system analysis code, TRACE, is chosen in this work. Its validation for stability analysis and comparison with the frequency domain approach, have been performed against the Ringhals 1 stability tests. Comprehensive assessment of modeling choices on TRACE stability analysis has been made, including effects of timespatial discretization, numerical schemes, thermal-hydraulic channel grouping, neutronics modeling, and control system modeling. The predictions from both the TRACE and STAB codes are found in reasonably good agreement with the Ringhals 1 test results. The biases for the predicted global decay ratio are about 0.07 in TRACE results, and -0.04 in STAB results. However, the standard deviations of decay ratios are both large, around 0.1, indicating large uncertainties in both analyses. Although the TRACE code uses more sophisticated neutronic and thermal hydraulic models, the modeling uncertainty is not less than that of the STAB code. The benchmark results of both codes for the Ringhals stability test are at the same level of accuracy. The fuel cladding integrity during power oscillations without reactor scram is examined by using the FRAPTRAN code, with consideration of both the stress-strain criterion and thermal fatigue. Under the assumed power oscillation conditions for high burn-up fuel, the cladding can satisfy the stress-strain criteria in the ASME Code. Also, the equivalent alternating stress is below the fatigue threshold stress, thus the fatigue limit is not violated. It can be concluded that under a large amount of the undamped power oscillation cycles, the cladding would not fail, and the fuel integrity is not compromised. / by Rui Hu. / Ph.D.

Thesis (S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2008. / "April 2008." / Includes bibliographical references (p. 144-147). / Analysis has been undertaken to investigate the utilization of fuel assembly venting in the reference design of the gas-cooled fast reactor under study as part of the larger research effort at MIT under Gen-IV NERI Project No. 05-044, "Optimized Competitive Supercritical-CO₂ Cycle GFR for Gen-IV Service." The focus of this work is on selection and optimization of a fuel assembly configuration best suited for venting, assessment of the radionuclide release of such an assembly design, and identification of plant systems which must be altered in order to support fuel venting. The innovative tube-in-duct design of the reference GFR fuel assembly is particularly well suited to venting, allowing fission products to diffuse into a common header before being routed along a vent path and eventually being released to the primary coolant system. A set of equations were developed which model the transport of fission products from fuel to vent path to primary coolant inventory and then into the containment atmosphere, with emphasis on conservatism in calculations of coolant impurity levels. Using these equations in a computer code, the lengthy list of radioactive and chemically volatile fission products for study was reduced to only fifteen species of any great concern. Of these, 85Kr and 137Cs were considered conservative bounding cases on the behavior of the other nuclides. The chemistry of the fission products released to the coolant was explored. In particular, reactions between fission products and their surroundings were identified, and estimates of deposition of both compounds and free atoms on steel surfaces were made. Investigation of reactions between fission products and CO₂ revealed that the formation of stable oxides would lead to deposition of most volatile species within the fuel assembly's debris trap. / (cont.) The radioactive volatiles which remain in the primary comprise only tritium, two isotopes of iodine, two isotopes of tellurium, and three isotopes of cesium. These species deposit on primary surfaces to a great enough degree to preclude hands-on primary system maintenance. In addition, techniques for removal of volatile and radioactive species from the primary coolant were identified, and a scheme for purification of the primary coolant volume was selected. The analysis confirms that primary coolant activity can be maintained at acceptable levels when purification of the primary coolant volume occurs once per hour. The response of the vent system to changes in primary system pressure was also investigated. In particular, the small periodic pressure transients known as breathing were studied, as were loss-of-coolant type scenarios. Both cases were investigated using hand calculations and computational techniques, and the radiological consequences of each were addressed. The analysis demonstrated the ability of the vented fuel assembly to quickly equalize pressure in the event of a sudden drop in operating pressure, and also showed the ability of a succession of plena to prevent the expulsion of much of the activity inventory of the plena during a breathing transient. Conversely, the use of a single plenum results in reduced activity release during a LOCA. After consideration of both results, a two-plenum approach, with a large lower annular plenum following a small upper axial plenum, was selected as the final design. / by Stephanie A. McKee. / S.M.

Thesis (S.B.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2007. / "June 2007." / Includes bibliographical references (leaf 44). / In the transportation sector, the current dependence on petroleum to satisfy large transportation fuel demand in the US is unsustainable. Oil resources are finite, and causing heavy US reliance on oil imports. Therefore, the development of alternative transportation fuels that do not depend on oil is becoming increasingly necessary. Our research investigates the feasibility of producing gasoline synthetically from nuclear hydrogen and two carbon sources: carbon dioxide emissions and municipal solid waste. These synthetic fuels have the potential to satisfy the large demand for gasoline, while reducing CO2 emissions. The nuclear hydrogen is produced through High Temperature Steam Electrolysis (HTSE), with heat and electricity provided by a supercritical CO2 cooled gas fast reactor. Through this study, we determine the suitable components for gasoline production from CO2 emissions and MSW. The feasibility of these methods of gasoline production was assessed by performing material and energy balances for the involved processes, determining preliminary cost estimates, and evaluating production scale and environmental impact. The material balances were compatible with our gasoline production scheme. By-product oxygen from the HTSE was especially beneficial for both production schemes, leading to various efficiency improvements. Water that is generated in the production processes can also displace a portion of water input for HTSE. By matching HTSE H2 output with H2 requirement of each production scheme, gasoline can be produced on a large scale. Gasoline output from MSW and coal plant CO2 emissions was about 1 million gallons/day and 550,000 gallons/day, respectively. / (cont.) These gasoline outputs are similar to SASOL Fischer-Tropsch plant in South Africa and the New Zealand methanol-to-gasoline plant. The base price of our synthetic gasoline was $4.35/gallon and $4.04/gallon for gasoline produced from CO2 and MSW, respectively. These costs will not be competitive with current US oil prices, but has high potential to compete with unconventional oil sources if oil prices rise significantly in the future. Carbon dioxide emissions can be significantly reduced with both production schemes, with MSW producing zero net emissions. / by MinWah Leung. / S.B.

Thesis (S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2008. / Cataloged from PDF version of thesis. / Includes bibliographical references. / The Department of Energy recently submitted a license application for the Yucca Mountain repository to the Nuclear Regulatory Commission, yet even the most optimistic timetable projects that the repository will not now open until at least 2020. The Office of Civilian Radioactive Waste Management recently revised the official undiscounted total life cycle cost of the waste management system upward by $22B (2000$), an increase of nearly 40% over the previous estimate, published in 2001. In this thesis a waste management tool, named SNuFManager (Spent Nuclear Fuel Manager), has been developed which deterministically simulates the stocks and flows of spent fuel in the United States and estimates annual expenditures based on the system's behavior. The tool allows policy makers to quickly and cheaply estimate the economic consequences of various decision alternatives under an array of scenarios in order to make quantitatively informed decisions and identify ways to mitigate or reverse recent increases in life cycle costs. The results are expressed in 2000 dollars, enabling a convenient comparison with the government's 2001 total system life cycle cost analysis. For each year of delay beyond 2020 in opening the repository and transferring ownership of spent fuel to the federal government, the total waste management system life cycle cost is estimated to increase by another $330M (2000$). The model also estimates that switching from the current mined geologic repository approach to a deep borehole disposal strategy would reduce the undiscounted total system life cycle cost by $19.4B, or 32%. / (cont.) Assuming a 10% discount rate, the net present cost of the deep borehole strategy is 18% less than that of the mined geologic repository approach. Finally, the model illustrates the economic benefits of opening a centralized interim storage facility of significant capacity as soon as possible. For example, if a 40,000 metric tonne facility, comparable in scale to the proposed Private Fuel Storage Facility in Utah, was opened by 2020, and the mined repository was opened in the same year, the total life cycle cost would be reduced by $1.5B relative to the case with no interim storage. If, moreover, the opening date of the mined geologic repository were delayed until 2040 or 2060, the savings provided by interim storage increase dramatically, to $4.9 and $8.1B, respectively. The thesis concludes with a discussion of the political and strategic consequences of several key policy choices. / by Taylor Allen Moulton. / S.M.