Thesis (S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2013. / Cataloged from PDF version of thesis. / Includes bibliographical references (page [43]). / Testing was performed on a data acquisition (DAQ) system that was built specifically to characterize a new ultracompact superconducting isochronous cyclotron (USIC) at MIT. A production model of Nitrogen-13 was validated by experiment at Massachusetts General Hospital. Both absolute and relative activity calculations were conducted utilizing an original MATLAB script; it is demonstrated that relative activity calculations are both more accurate and precise. The DAQ system limitations in terms of potential activity calculations were also explored. A saturation limit of 3500 coincident counts per second was calculated and validated, which sets a design criterion limit for future work. The potential for the USIC to be utilized in an active SNM interrogation role is also explored. / by Shawn Fitzgerald. / S.M.

Thesis (S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2008. / Page 128 blank. Cataloged from PDF version of thesis. / Includes bibliographical references (p. 50-52). / A comprehensive risk-informed methodology for passive safety system design and performance assessment is presented and demonstrated on the Flexible Conversion Ratio Reactor (FCRR). First, the methodology provides a framework for risk-informed design decisions and as an example two design options for a decay heat removal system are assessed and quantitatively compared. Next, the reliability of the system is assessed by quantifying the uncertainties related to system performance and propagating these uncertainties through a response surface using Monte Carlo simulation. Finally, a sensitivity study is performed to measure the relative effects of each parameter and to identify ways to maintain, improve, and monitor system performance. / by Christopher J. Fong. / S.M.

Thesis (S.B.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2010. / Cataloged from PDF version of thesis. / Includes bibliographical references (p. 56). / A study on the dynamics of filaments in the presence of a diagnostic, conductive limiter is presented. Plasma filaments are coherent structures present in many fusion devices and transport a significant amount of particles and energy to vacuum chamber walls. The theories of filament propagation are based on a circuit model of current closure loops within the filament and are described herein. Two experimental configurations based on different modes of filament generation are utilized in the Versatile Toroidal Facility (VTF) to test the model. Along with the experimental observations of filament propagation, measurements of limiter sheath resistance are made and shown to depend upon environmental conditions and filament parameters. One experimental configuration creates long filaments (-6 meters) using electron cyclotron resonance heating (ECRH) and a toroidally symmetric solenoid. The other utilizes a newly constructed Argon plasma gun breaking down injected gas for creating short (~1 meter) plasma filaments. Arrays of Langmuir probes track the path of the plasma particles while the limiters are constructed to generate a uniform vertical electric field along its conductive side and measure the parallel current collected. Agreement is found with the model in a region of strongly negative vertical electric fields. Current collection data extends agreement to larger regions with eventual breakdown at strongly positive electric fields due to complex current collection and escape paths. / by Jeffrey David Bonde. / S.B.

Thesis (S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2008. / Includes bibliographical references (leaves 66-67). / Nanofluids, colloidal dispersions of nanoparticles in a base fluid, have shown enhancements in both pool boiling and flow boiling critical heat flux (CHF) in laboratory tests. The applicability of this aspect to nuclear reactor post-accident cooling is promising. This study investigates various parameters that should be considered for such applications. Dilute alumina nanofluids were tested in terms of radiation durability, chemical stability, and CHF enhancement capability in wire and plate geometries. They have been found to be stable under high doses of gamma radiation and in the chemical environment associated with emergency core cooling. However, exposure to tri-sodium phosphate, a chemical used in in-vessel retention systems, resulted in significant agglomeration of nanoparticles. CHF value increases were obtained for both nanofluid and primary coolant chemistry conditions. Additionally radiation induced surface activation (RISA) effect on CHF was evaluated for alumina and titania nanoparticle depositions on plate geometries. Pool Boiling experiments were conducted with both wire and plate heaters. Wire test results showed CHF enhancement of 23% to 74% for dilute alumina nanofluids of 0.001v% to 0.1v%, with the highest enhancement obtained for lowest concentration alumina nanofluid. With the addition of boric acid in alumina nanofluid, the highest CHF enhancement is up to 90%. RISA was proven to play a significant role in CHF increase. Stainless steel plates pre-coated with titania and alumina nanoparticles showed CHF enhancement of 146% and 133%, respectively, compared to CHF obtained from plain heaters, after irradiation in a Co-60 gamma source. These initial results reveal that applying nanofluid in reactor chemistry and radiation environment can potentially bring about significant benefit in increasing the safety margin. Further study is needed to elucidate these phenomena in prototypic flow boiling, chemistry, and radiation conditions. / by Timothy R. Lucas. / S.M.

Thesis (S.B.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2007. / Includes bibliographical references (leaves 45-50). / Carbon dioxide is suspected to be a major contributor to global warming. In the United States, nearly 70% of electricity is produced using coal or natural gas, both of which emit carbon dioxide into the environment. Nuclear power, which does not emit any carbon dioxide, produces 17% of the electricity consumed in the United States. In order to persuade utilities to switch from coal or natural gas to nuclear power and thus reduce carbon dioxide emissions, a carbon tax should be implemented. Depending on the cost of construction for new nuclear plants and the level of savings that will incentivize utilities to switch, the carbon tax needed to promote nuclear power will range between $20/tC and $200/tC. The full range of carbon tax scenarios are developed in this thesis, with the most likely carbon tax being $1 10/tC. This cost assumes a $1800/kw capital construction cost and a 10% risk perception premium on the bus bar cost of power to address the financial and industry community's somewhat negative perception of nuclear investments. From a policy perspective, this carbon tax will be more effective in causing utilities to move to nuclear power than a cap and trade policy. From a utility standpoint, switching to nuclear power under a carbon tax is less expensive than implementing a program of carbon capture and sequestration. / by Katherine C. Thornton. / S.B.

Thesis: S.M. and S.B., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 2015. / Cataloged from PDF version of thesis. / Includes bibliographical references (pages 135-138). / This thesis details the ongoing development of a new Offshore Floating Nuclear Plant (OFNP) concept that exhibits a promising potential for economic and rapid deployment on a global scale. The OFNP creatively combines state-of-the-art Light Water Reactors (LWRs) and floating platforms similar to those used in offshore oil and gas operations. A reliable and cost-effective global supply chain exists for both technologies, which enables a robust expansion in the use of nuclear energy on a time scale consistent with combating climate change in the near future. The OFNP is a plant that can be entirely built within a floating platform in a shipyard, transferred to the site, where it is anchored within 12 nautical miles (22 km) off the coast in relatively deep water (=/> 100 m), and connected to the grid via submarine AC transmission cables. Shipyard construction ensures a supply of qualified workers and facilities, and it brings mass-production-like construction efficiency to existing reactor designs. Eventual shipyard decommissioning allows sites to immediately return to a "green field" condition when the plant's life is spent. The crews would operate in monthly or semi-monthly shifts with onboard living quarters, similar to oil and gas platforms. The OFNP is a nuclear plant specifically designed for the global market: it can be constructed in one country or multiple countries and exported internationally. It lends itself to a flexible and mobile electricity generation strategy, which minimizes the need for indigenous nuclear infrastructure in the host country and does not commit the customer to a 40 to 60 years-long project. / by Jacob M. Jurewicz. / S.M. and S.B.

Thesis: Ph. D., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 2015. / Cataloged from PDF version of thesis. / Includes bibliographical references (pages 294-305). / The Fluoride Salt-Cooled High-Temperature Reactor (FHR) is a pebble bed nuclear reactor concept fueled by tristructural isotropic (TRISO) fuel particles embedded in graphite spheres and cooled by a liquid fluoride salt known as "flibe" (7LiF-BeF2). A system of models was developed which enabled analyses of the performance of a prototypical pebble bed FHR (PB-FHR) with respect to tritium production and transport, corrosion, TRISO fuel performance, and materials stability during both normal and beyond design-basis accident (BDBA) conditions. A model of TRITium Diffusion EvolutioN and Transport (TRIDENT) was developed and benchmarked with experimental data. TRIDENT integrates the effects of the chemical redox potential, tritium mass transfer, tritium diffusion through pipe walls, and selective Cr attack by tritium fluoride. Systems for capturing tritium from the coolant were proposed and simulated with TRIDENT. A large nickel permeation window reduced the tritium release rate from 2410 to 800 Ci/EFPD. A large gas stripping system reduced tritium release rates from 2410 to 439 Ci/EFPD. A packed bed of graphite located between the reactor core and the heat exchanger reduced peak tritium release rates from 2410 to 7.5 Ci/EFPD. Increasing the Li-7 enrichment in flibe from 99.995 to 99.999 wt% reduced both the tritium production rate and the necessary sizes of tritium capture systems by a factor of 4. An existing TRISO fuel performance model called TIMCOAT was modified for use with PBFHRs. Low failure rates are predicted for modern uranium oxycarbide (UCO) TRISO fuels in a PBFHR environment. Post-irradiation examinations of surrogate TRISO particles determined that the outer pyrolytic carbon layer is susceptible to cracking if flibe were to freeze around the particles. Chemical thermodynamics calculations demonstrated that common constituents of concrete will not be stable in the event they contact liquid flibe. The chemical stability of fission products in reference to the coolant redox potential was determined in the event the TRISO UCO kernel is exposed to flibe during a BDBA. Noble gases (Kr and Xe) will escape the coolant. Cesium, strontium, and iodine are retained in the salt. All other important radionuclides are retained in the kernel or within the coolant system. / by John Dennis Stempien. / Ph. D.

Thesis (S.M. and S.B.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2008. / Page no. "2" used twice. Cataloged from PDF version of thesis. / Includes bibliographical references (p. 147-150 [i.e. p. 148-151]). / The MIT Nuclear Research Reactor (MITR) is the only research reactor in the United States that utilizes plate-type fuel elements with longitudinal fins to augment heat transfer. Recent studies on the conversion to low-enriched uranium (LEU) fuel at the MITR, together with the supporting thermal hydraulic analyses, propose different fuel element designs for optimization of thermal hydraulic performance of the LEU core. Since proposed fuel design has a smaller coolant channel height than the existing HEU fuel, the friction pressure drop is required to be verified experimentally. The objectives of this study are to measure the friction coefficient in both laminar and turbulent flow regions, and to develop empirical correlations for the finned rectangular coolant channels for the safety analysis of the MITR. A friction pressure drop experiment is set-up at the MIT Nuclear Reactor Laboratory, where static differential pressure is measured for both flat and finned coolant channels of various channel heights. Experiment data show that the Darcy friction factors for laminar flow in finned rectangular channels are in good agreement with the existing correlation if a pseudo-smooth equivalent hydraulic diameter is considered; whereas a new friction factor correlation is proposed for the friction factors for turbulent flow. Additionally, a model is developed to calculate the primary flow distribution in the reactor core for transitional core configuration with various combinations of HEU and LEU fuel elements. / by Susanna Yuen-Ting Wong. / S.M.and S.B.

Thesis (S.M. and S.B.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering; and, (S.B.)--Massachusetts Institute of Technology, Dept. of Physics, 2012. / Cataloged from PDF version of thesis. / Includes bibliographical references (p. 105-107). / In this thesis, I designed and implemented a particle-in-cell (PIC) code on a graphical processing unit (GPU) using NVIDA's Compute Unified Architecture (CUDA). The massively parallel nature of computing on a GPU nessecitated the development of new methods for various steps of the PIC method. I investigated different algorithms and data structures used in the past for GPU PIC codes, as well as developed some of new ones. The results of this research and development were used to implement an efficient multi-GPU version of the 3D3v PIC code SCEPTIC3D. The performance of the SCEPTIC3DGPU code was evaluated and compared to that of the CPU version on two different systems. For test cases with a moderate number of particles per cell, the GPU version of the code was 71x faster than the system with a newer processor, and 160x faster than the older system. These results indicate that SCEPTIC3DCPU can run problems on a modest workstation that previously would have required a large cluster. / by Joshua Estes Payne. / S.B. / S.M.and S.B.

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 212-221). / Fluoride Salt Cooled High Temperature Reactors (FHRs) are a new reactor concept that have recently garnered interest because of their potential to serve missions and generate revenue from sources beyond those of traditional base-load light water reactor (LWR) designs. This potential is facilitated by high-temperature, atmospheric-pressure operation enabled by the incorporation of liquid fluoride salt coolants together with solid microparticle TRISO fuel. Since no FHR has been built, an important technology development step is the design, construction, and operation of a FHR test reactor (FHTR). The FHTR's strategic goals cannot be satisfied using small-scale experiments or test loops: (1) develop the safety and licensing basis for a commercial plant; (2) demonstrate technological viability and provide operational and maintenance experience; and (3) test alternative fuels, fluoride salt coolants, and structures in an actual reactor configuration. The goals of the FHTR support the development of the commercial FHR, but are different. The programmatic goals for the FHTR drive the specification of the technical design goals: (1) capability to switch between any one of various potential liquid fluoride salt coolants; (2) provide an irradiation facility for accelerated fuels and materials testing. The first stage of the present work included an exploration and characterization of the available design space for an FHTR. Many different core, reflector, and assembly designs were evaluated to determine configurations that possessed acceptable performance while satisfying all design constraints. This work resulted in a novel prismatic block assembly design termed Fuel Inside Radial Moderator (FIRM), which leverages spatial selfshielding of the fuel microparticles to increase core reactivity by ~10,000 pcm relative to a traditional prismatic block design, enabling operation with any of the proposed liquid fluoride salt coolants. This stage of work served to focus the search space for the application of formal optimization algorithms to further improve the feasible design. The second stage of the present work involved the development of a methodology to perform full-core optimization of the feasible FHTR design and its implementation into usable software. The OpenFRO (Open source Framework for Reactor Optimization) code implements the Efficient Global Optimization (EGO) surrogate-based optimization framework, which has been successfully applied to aerospace and automotive engineering optimization problems in the past. OpenFRO extends the EGO framework to full-core reactor optimization in the presence of uncertainty, enabling an effective, automated, and efficient approach for earlystage reactor design. OpenFRO's EGO implementation imposes minimal computational overhead while reducing the number of required high-fidelity simulations for optimization by 96%. The final stage of the present work involved the identification and analysis of the optimal design of the FHTR. The optimal design was selected based on its capability to provide the best performance across potential salt coolants and power levels. The optimal design achieved irradiation position fluxes 90%-130% greater than the feasible design initially identified, while satisfying all safety and performance constraints. / by Joshua Richard. / Ph. D.