Thesis: S.M., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 2015. / Cataloged from PDF version of thesis. / Includes bibliographical references (page 82). / Conduction and single-phase convective heat transfer are well understood phenomena: analytical models [1] and empirical correlations [2] allow capturing the thermal behavior of plate-type fuels or heaters in contact with a single-phase coolant. On the other hand, transient boiling heat transfer is a scarcely studied and much less understood phenomenon. Although, earlier studies have shown that important features of the boiling curve (i.e. onset of nucleate boiling (ONB), nucleate boiling heat transfer coefficient, and critical heat flux (CHF)) in transient conditions. These parameters significantly differ from those at steady-state. The mechanisms by which these changes occur are not clear. Furthermore, some of the conclusions from different authors are quantitatively or qualitatively in disagreement with each other. This work studied transient pool boiling heat transfer phenomena under exponentially escalating heat fluxes on plate-type heaters, at the time scales of milliseconds typical of Reactivity Initiated Accidents (RIAs) in nuclear reactors. The investigation utilized state-of-the-art diagnostics such as Infrared (IR) thermometry and high-speed video (HSV), to gain insight into the physical phenomena and generate a database that could be used for development and validation of accurate models for transient boiling heat transfer. The tests with exponential power escalation periods ranging from 100 ms to 5 ms and subcoolings of OK (saturation), 25K and 75 K were conducted. The measured pre-ONB heat transfer coefficient agrees well with the theoretical predictions for transient conduction. The ONB and onset of significant void (OSV) temperature and heat flux were found to increase monotonically with decreasing period and increasing subcooling, as expected. The mechanistic ONB model of Hsu was able to predict the measured ONB temperature and heat flux. The transient pool boiling curves were measured up to fully developed nucleate boiling (FDNB). Generally two types of boiling curve were observed: with overshoot (OV) or without overshoot. Data show that, when an OV is present, the OV temperature increases monotonically with decreasing period and increasing subcooling. The present study clears the confusions (eg. the trend of ONB temperature and heat flux versus power period) in previous research, and sheds light to the mechanisms behind transient boiling heat transfer. This can ultimately reduce the uncertainty in both design and safety analyses of the research reactors especially under RIAs. / by Guanyu Su. / S.M.

Thesis: Ph. D., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 2016. / 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 183-196). / Hydrogen embrittlement of engineering alloys is characterized by a loss of ductility and unpredictable failure. These failures affect numerous industries, including nuclear power, oil and gas exploration, and hydrogen transportation and storage. In face-centered cubic alloys, the resultant fracture is intergranular and very sensitive to grain boundary character. We study this behavior in alloy 725, a popular nickel-iron superalloy with high strength and corrosion resistance. Using a suite of complementary experimental techniques we reveal the fracture behavior of individual grain boundaries in hydrogen embrittlement for the first time, providing critical understanding of the role of grain boundary character and informing improved microstructure design. We study crack propagation in hydrogen embrittled tensile test specimens using highenergy diffraction-microscopy, a non-destructive X-ray synchrotron technique capable of mapping grain boundaries in 3-D. We find that boundaries with low-index planes (BLIPs), defined as planes within 10° of [111], [110] or [100], resist crack propagation and improve toughness. We show that coherent twin boundaries (CTBs), a subset of BLIPs, also indirectly improve toughness by increasing the heterogeneity of the grain boundaries they intersect. In addition, we use electron backscatter diffraction and scanning electron microscopy to identify the grain boundaries along which cracks initiate and propagate on the sample surface. We unambiguously show that grain boundaries are the source of crack initiation, and we study a statistically significant number of cracking events, providing the ability to determine the role of grain boundary character. Surprisingly, we find that while CTBs resist crack propagation, they preferentially initiate cracks. These results inform a more nuanced approach to microstructure design. Typically grain boundary engineering techniques aim to maximize the fraction of low-S boundaries as designated by the coincident site lattice model. Our results suggest that these techniques should maximize the fraction of BLIPs instead. In addition, the dual nature of CTBs suggests the development of graded microstructures, with high concentrations of CTBs in the interior to resist crack propagation and reduced concentrations at the surface to limit crack initiation. / by John Paul Hanson. / Ph. D.

Thesis (S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2010. / Cataloged from PDF version of thesis. / Includes bibliographical references (p. 83-85). / Alcator C-Mod is a deuterium tokamak reactor experiment operated by the MIT Plasma Science and Fusion Center. Following the 2008 Alcator C-Mod campaign, the reactor was shut down and opened for maintenance and upgrades. During this time, it was discovered that the boronized TZM tiles, with boron films applied in situ, lining the inner surfaces of the reactor had experienced non-uniform damage as a result of plasma interactions. This damage was most pronounced in the tiles from the limiter and divertor, but also appeared in regions of the reactor that were not traditionally believed to experience high heat or particle fluxes. The objective of this thesis is to perform a thorough microstructural analysis of these tiles, particularly the boron-TZM interface, in order to explore the damage mechanisms present. In turn, the details of the damage mechanisms will illuminate the plasma parameters, such as temperature and particle flux, that caused the damage. This analysis will also allow for a prediction of the behavior of other tile and coating materials under consideration for use in fusion devices. During this investigation, a number of tiles with varying degrees of apparent damage were removed from the limiter and divertor and photographed for macroscopic characterization. Modeling was also performed using the expected heat and particle fluxes at the tile surface, along with the thermal transient history of the tiles, to estimate the depth of damage from each of these sources. These results were compared to data gathered during analysis of the tiles through several techniques, including ion beam analysis, scanning electron and optical microscopy, energy dispersive X-ray spectroscopy, and X-ray diffraction spectroscopy. Ultimately, the diverse sets of data gathered through these techniques provided for a fairly cohesive rationalization of the damage mechanisms present. In the case of the divertor tiles, no thermal damage was observed, but the boron film was eroded through sputtering as a result of the large fluxes of high energy particles encountered in that region of the reactor. Tiles in the limiter, however, experienced a more severe amount of damage caused primarily by thermal effects. The surface temperatures at these tiles were in the range 2140-2600 degrees Celsius, surpassing the melting point of boron and approaching that of the underlying TZM. Recrystallization of the TZM substrate was observed to an average depth of -20m, with an overall observed heat penetration depth of -100m. These temperatures indicate a local heat flux of ~108- 109W/m 2 when applied to the heat diffusion model used earlier in this investigation. Such a large heat flux indicates a transient event responsible for the observed damage, occurring on a timescale of milliseconds rather than the one second duration of a pulse at peak power. This transient would be characterized by an increase by an order of magnitude of the product of the plasma density and sheath temperature. Another possibility is that the beads which formed on the melted surface extended beyond the plasma Debye length and intersected the magnetic field lines, resulting in an increase in the heat fluxy by an order of magnitude at those locations. It is difficult to separate the contributions of these damage mechanisms from data obtained after a full Alcator campaign, and further investigation is warranted to better understand each of these processes. / by Joseph Michael Hubley. / S.M.

Thesis (S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2009. / Cataloged from PDF version of thesis. / Includes bibliographical references (p. 143-144). / Compared to the traditional solid fuel geometry for PWRs, the internally and externally cooled annular fuel offers the potential to increase the core power density while maintaining or increasing safety margins. It is demonstrated that for the Korean OPR-1000 reactor, power density can be increased by 20% when the 16x16 solid fuel assemblies are replaced by 12x12 annular fuel assemblies. In this annular fuel design, the assembly dimensions, coolant flow rate, and core outlet coolant temperature are kept fixed at the reference values for the OPR-1000 with solid fuel. The core inlet temperature is decreased to accommodate the additional 20% energy. Thermal hydraulic steady state analyses are carried out to determine the Minimum Departure Nucleate Boiling Ratio (MDNBR) margin and evaluate improvement in the design to maximize this margin. Whole core VIPRE-01 model results show that a proposed 14x14 annular fuel design cannot achieve high power uprate because of sub-limit MDNBR in the inner channel. To better optimize the 12x12 annular fuel design, the rod dimensions are fine-tuned by slightly increasing the inner channel diameter and outer channel diameter, while keeping the fuel to moderator ratio fixed. The modified design can achieve 20% power uprate. In addition, MDNBR sensitivity to manufacturing tolerances is investigated, showing that the new proposed design can accommodate typical manufacturing tolerances. Partial blockage at the inlet of the inner channel and the impact of corrosion and crud growth are also analyzed by conservertive models. The inner channel can accommodate a blockage of up to 43% of its flow area before MNDBR falls below the 1.3 limit.The crud and ZrO2 buildup does not reduce MDNBR margin below the 1.3 limit, as long as the combined thickness is less than 74[mu]m-94[mu]m. Neutronic analyses are performed for OPR-1000 with both the solid fuel and the annular fuel. The results from an MCNP model of the reference solid fuel assembly and a CASMO-4 model show excellent agreement. The benchmark of annular fuel array shows that CASMO-4 overpredicts the eigenvalues and the slope of the reactivity burnup curve. Fictitiously increasing U-238 number densities in CASMO-4 inputs by 10% produces good match with the MCNP-based burnup code, MCODE2.2. The whole core model of Ulchin Nuclear Unit 5 is established as a benchmark using SIMULATE-3 to calculate the steady state reactor core performance. Last but not least, an equilibrium annular fuel core is proposed, and its steady state core performance is analyzed. The proposed annular fuel assemblies composed of 7.5% and 6.5% U-235 enriched fuel rods, and burnable poisons with various Gd 20 3 weight percentages (4%, 6%, 8%, 10%, and 16%) can satisfy the design targets, such as peak boron concentration, cycle length, and peaking factors in a certain equilibrium loading pattern. / by Liang Zhang. / S.M.

Thesis (S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2006. / This electronic version was submitted by the student author. The certified thesis is available in the Institute Archives and Special Collections. / Vita. / Includes bibliographical references (leaves 115-117). / A helicon plasma source experiment has been developed and then constructed in the MIT Space Propulsion Laboratory (SPL) vacuum chamber. This experiment allows study of the intrinsic advantages of efficient helicon plasma production for use in space electric propulsion. Historical helicon experiment data have been applied to help size the experiment. The goal was to create a robust and flexible experimental system which would allow optimization of the source and system parameters for efficient thrust generation, and would permit for correlation between helicon theory and experiment. This effort may lead to the development of a new electric propulsion device, the Mini-Helicon Plasma Thruster. A design process was undertaken for the creation of this experimental setup, with the aid of a number of students and researchers at the MIT SPL. This thesis will focus on the author's specific contributions to this larger effort, which included the following elements. A survey of past helicon experiment parameters was completed, made possible largely due to the wealth of data provided by helicon plasma use in academia and for research and development. An analysis of the flow of propellant through the thruster was completed, utilizing low Mach number flow theory. A metallic structure was designed, structurally analyzed and constructed to support the electromagnets used to provide the required magnetic field. In addition, a radio frequency matching network enclosure and suitable interconnections were designed and constructed as part of the RF power delivery system. The result of the design and construction effort is a working, reliable, and flexible helicon plasma source system. This system provides the capability for future experimentation and helicon plasma thruster development. / by Joseph Eugene Palaia, IV. / S.M.

Thesis (S.M. and S.B.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2007. / Includes bibliographical references (leaves 106-107). / The objective of this work was to measure the effective thermal conductivity of a number of materials (particle beds, and fluids) proposed for use in and around canisters for disposal of high level nuclear waste in deep boreholes. This information is required to insure that waste temperatures will not exceed tolerable limits. Such experimental verification is essential because analytical models and empirical correlations can not accurately predict effective thermal conductivities for complex configurations of poorly characterized media, such as beds of irregular particles of mixed sizes. The experimental apparatus consisted of a 2.54 cm. diameter cylindrical heater (heated length = 0.5 m) , surrounded by a 5.0 cm inner diameter steel tube. Six pairs of thermocouples were located axially on the inside of the heater sheath, and in grooves on the air-fan-cooled outer tube. Test media were used to fill the annular gap, and the temperature drop across the gap measured at several power levels covering the range of heat fluxes expected on a waste canister soon after emplacement. Values of effective thermal conductivity were measured for air, water; particle beds of sand, SiC, graphite and aluminum; and an air gap subdivided by a thin metal sleeve insert. Results are compared to literature values and analytical models for conduction, convection and radiation. Agreement within a factor of 2 was common, and the results confirm the adequacy, and reduce the uncertainty of prior borehole system design calculations. All particle bed data fell between 0.3 and 0.5 W/moC, hence other attributes can determine usage. / by Samina Shaikh. / S.M.and S.B.

Thesis (S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2007. / Includes bibliographical references (p. 129-130). / Comparing different nuclear fuel cycles and assessing their implications require a fuel cycle simulation model as complete and realistic as possible. In this thesis, methodological implications of modeling choices are discussed in connection with development of the MIT fuel cycle simulation code CAFCA. The CAFCA code is meant to find the recycling facilities deployment rate that minimizes the time by which spent fuel in storage today is used up in order to lead to a nuclear fuel cycle with minimum inventory of transuranic elements. The deployment is constrained by the recycling plants construction capacity and also by the economic requirement that the recycling plants loading factor never drops below a certain level. First, through a simplified fuel cycle model, it is analytically proven that an optimum solution is to build recycling plants at full speed up to a certain point in time b, then to suspend construction until interim storage gets completely depleted. The shape of the optimum solution, parameterized by b, is injected into an algorithm based on a complete model of the fuel cycle. An iterative process yields the value of b assuring depletion and satisfactory loading factors. Besides providing rigorous optimization,this analytical solution underpinning the CAFCA algorithm is expected to reduce considerably the vulnerability of the results to numerical discontinuities. Degradation of fuel quality with time in interim storage occurs due to the decay of Pu241 into Am241. While an obvious approach to track such effects is to couple the fuel cycle code with a neutronics/decay code (ORIGEN for example), it is more efficient to derive explicit equations from a simplified irradiation and decay model, allowing for analytical tracking of the fuel composition. / (cont.) All fuel cycle simulation refinements do not present the same level of importance. One should focus on the dominant parameters as those contributing the most to overall results sensitivity. A novel U.S. thermal recycling scenario called CONFU is taken as a reference case. The CONFU technology is introduced 15 years from now, with an industrial capacity allowing the construction of one 1000 MT/year spent fuel separation plant every two years. Discharged CONFU batches remain in cooling storage for 6 years. Reactors have a 60 years lifetime and economic ecovery period of 20 years, and are half financed by equity with a rate of return of 15%. It is found that the cost of electricity is most sensitive to the reactors lifetime, since taking it back to its initial nominal value of 40 years would result in a 44% increase in the cost of electricity. Next in importance is the financing structure of the fleet. The addition of three points to the rate of return on equity would increase the cost of electricity by 14%. While scale effects are locally very beneficial in that they substantially reduce recycling plants operation costs, they prove to be of limited interest from an overall fuel cycle point of view. Using the scale effect model in CAFCA-II, doubling the separation plants capacity yields a 3% reduction of the cost of electricity. The fuel cycle presents good robustness with respect to fuel decay time degradation. Increasing CONFU batches cooling time to 18 years causes a 2% increase in the cost of electricity. / by Nicéphore Bonnet. / S.M.

Thesis: S.M., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 2016. / Cataloged from PDF version of thesis. / Includes bibliographical references (pages 73-76). / In most tokamak fusion reactor designs, ICRF (Ion Cyclotron Range of Frequencies) and LH (Lower Hybrid) radio frequency (RF) waves used to heat the plasma and drive current are launched from the low-field, outboard side where there is more access space. It has recently been proposed to launch these waves from the high-field side [1-3], which increases current-drive efficiency, allows for better wave penetration, and has favorable scrape-off-layer and plasma material interaction characteristics [4]. However the poloidal location and size of RF launchers will also affect important aspects of the neutronics of the tokamak fusion design, i.e. how the 14.1 MeV neutrons born out of the deuterium-tritium (D-T) fusion reaction interact with the surrounding blanket and structures. The goal of this thesis is to assess the dependence of RF launcher poloidal location on the important neutronics parameters of tritium fuel breeding, launcher damage and activation. To determine the effects of waveguide and antenna location on Tritium Breeding Ratio (TBR), damage, and activation, the MCNP Transport Code was used, as well as the EASY 2010 activation package to analyze the activation of the vacuum vessel components. A simple geometry was designed for MCNP, based on the original ARC model [1]. Seven locations for the waveguides and antenna were chosen: the inner and outer midplane, the inner and outer upper corners, two spaces between the midplane (inboard and outboard), and a central location directly above the vacuum vessel. TBR, DPA, and helium concentration were calculated at all seven points to find the optimal location for the waveguides and antenna. Four blanket materials were chosen: two liquid blankets (FliBe and Pb-17Li) and two solid blankets (Li4SiO4 and Li2TiO3). This was to test whether or not blanket material affects the optimal location of the launchers. We find that from the neutronics point of view the overall optimal location is the inboard upper corner, which minimizes DPA and helium concentration in the antenna and waveguide, and maximizes TBR. DPA in the waveguide was minimized when placed in the outboard upper corner, although the difference in DPA between the two locations was small. While TBR was maximized at the top of the vacuum vessel, the differences in TBR between all locations was less than 1%. These results reinforce the choice of inside, upper corner launch as the optimal location for current drive, launcher protection and neutronics. Activation was also assessed for the vacuum vessel, both without and with the waveguides and antenna, assuming irradiation times of one week, one month, and one year. Overall, activation was significant in the vacuum vessel, as expected, due to the use of Inconel 718. The IAEA recycling limit could be achieved, regardless of irradiation time. The dominant isotopes present after irradiation differed when the irradiation time was one week versus one month or one year. Activation was also assessed in the waveguides and antenna for the cases of the launchers being placed at the outboard midplane versus the inboard corner. The activation in the antenna was shown to be reduced by a factor of two and in the waveguides by a factor of four, when the launchers were placed in the inboard corner. / by Jennifer Marie Sierchio. / S.M.

Thesis (S.M. and S.B.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2011. / In title on title-page, double underscored "m" appears superscript. Cataloged from PDF version of thesis. / Includes bibliographical references (p. 54-57). / Simultaneous ⁹⁹m̳̳Tc-MDP/ ¹²³I-MIBG SPECT has the potential to replace current clinical sequential acquisitions of ⁹⁹m̳Tc-MDP and ¹²³-MIBG SPECT studies, and therefore has great potential to reduce imaging time, sedation time, and injection dose on patients with neuroblastoma. Simultaneous ⁹⁹m̳Tc/¹²³ imaging is challenging because of the crosstalk between the ⁹⁹m̳Tc and ¹²³I photo-peak windows due to down-scatter of ¹²³I photons (159keV) to the ⁹⁹m̳Tc (140keV) photo-peak window and limited energy resolution of the scanner. Additionally, the counts of detected photons are limited because the injection dose as well as scan time are limited for neuroblastoma patients and scan acquisition cannot be performed for at least 24 hours after ¹²³I-MIBG injection. These factors hinder the separation of images of these two radionuclides. An enhanced fast Monte Carlo based joint ordered-subset expectation maximization (MC-JOSEM) reconstruction algorithm has been been developed for simultaneous ⁹⁹m̳Tc/¹²³ imaging by Ouyang, El Fakhri, and Moore (2007). MC-JOSEM incorporates attenuation into a full system matrix to greatly improve image accuracy and include both primary and scattered photons in the reconstruction process to significantly reduce image noise. Separate ⁹⁹m̳Tc-MDP and ¹²³I-MIBG Monte Carlo simulations were performed. For each isotope, noise-free projection data sets were generated first. Lesions were then added to the ⁹⁹m̳Tc and ¹²³I data separately. Mimicked dual-isotope data were then generated by combining the ⁹⁹m̳Tc and ¹²³I data. Images for single-isotope and dual-isotope were reconstructed by using standard clinical single-isotope OSEM and MC-JOSEM, respectively. Channel Hotelling observer (CHO) was used to calculate lesion detectability. On average, the CHO SNR obtained from dual-isotope studies is close to that of single-isotope studies for each radionuclide. (SNR: 3.521 for dual-isotope versus 3.828 for single-isotope). Hence, simultaneous ⁹⁹m̳Tc-MDP/ ¹²³-MIBG has the potential to replace sequential ⁹⁹m̳Tc-MDP and ¹²³I-MIBG for neuroblastoma imaging. / by José P. Pérez-Gutiérrez. / S.M.and S.B.

Thesis (S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2012. / Cataloged from PDF version of thesis. / Includes bibliographical references (p. 78-82). / The main goal of this project is to evaluate the feasibility of axial injection of a high brightness beam from an Electron Cyclotron Resonance ion source into a high magnetic field cyclotron. Axial injection from an ion source with high brightness is important to reduce particle losses in the first several turns of acceleration within the cyclotron. Beam brightness is a measure of the beam current and rate of spread of the beam. The ultimate goal in developing an ECR ion source is to enable reduced beam losses along the entire acceleration path from the ion source through the cyclotron, allowing for a high beam current accelerator. Cyclotrons with high beam current have the potential to improve the availability of proton radiation therapy. Proton radiation therapy is a precisely targeted treatment capable of providing an excellent non-invasive treatment option for tumors located deep within tissue. In order to model injection into high field it is necessary to measure the parameters of the beam extracted from the ion source. The two most important beam parameters are emittance and beam current. The emittance of the beam is a measurement of the rate of beam spread along the path of the beam and beam current is a measurement of the energy and quantity of particles within a charged particle beam. This thesis presents the design and analysis of an ECR Ion Source and the instruments used to measure the emittance and beam current. Based on the modeling of the ECR ion source beam and the data gathered during testing, the ECR ion source presented in this thesis has the potential to provide a high brightness beam capable of high field axial injection. Beam simulations provide insight into the performance of the ECR ion source in high magnetic field. Axial beam injection from an external ion source is promising with moderate refinements to the ECR ion source. / by Mark E. Artz. / S.M.