Thesis: S.B., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 2014. / "June 2014." Cataloged from PDF version of thesis. / Includes bibliographical references (page 22). / A significant effort is being made by fusion researchers to minimize the total size of magnetic fusion devices on the path toward developing fusion energy. The spherical tokamak, which has a very low aspect ratio, is the most promising of the compact magnetic fusion reactor designs. This compactness imposes a severe material constraint on the design, as a highly compact device will have very thin inner shielding. This inner shielding, which in traditional designs is required to be around 1 meter thick, acts to protect the central solenoid and return toroidal field coil legs from material damage and nuclear heating resulting from high neutron fluxes. The use of a sub-meter inner shield creates potential for the design of a proof of principle magnetic fusion device, sacrificing the central component materials for a demonstration of temporary fusion power production. The nuclear heating of thin shields (~ 0.1 - 0.2m) of various compositions was explored using the Monte Carlo N-Particle (MCNP) transport code. The principal finding was that nuclear heating is the largest concern to the central inboard components. Nuclear heating of these sensitive materials was found to be minimized by the use of a magnesium borohydride blanket with a tungsten first wall. The resulting nuclear heating density for a 100MW, R=1m D-T tokamak employing 0.1 - 0.2m shields is shown to have the potential to threaten the ability of such a device to sustain net electricity. / by Cameron T. French. / S.B.

Thesis (S.B.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2007. / "June 2007." / Includes bibliographical references (p. 40-41). / In order to study plasma rotation, temperature, and impurity density, a Neon Soft X-ray Spectrometer (NeSoXs) was installed on the Alcator C-Mod tokamak. This spectrometer used a spherically bent mica crystal as the reflective element in order to separate the spectral and spatial information on the receiving CCD. The original NeSoXs was found to have several problems including vacuum quality, camera readout rate, crystal positioning, chamber size, and chamber positioning. In order to remedy these problems, a new system was designed and has been constructed. This system has achieved much better vacuum pumping times (-100 times faster) and is in the process of being installed on the Alcator C-Mod vessel. / by Yuri Anatoly Podpaly. / S.B.

Thesis (S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2007. / Includes bibliographical references (p. 133-135). / The interaction between a probe and a plasma has been studied since the 1920s and the pioneering work of Mott-Smith and Langmuir [1], and is still today an active topic of experimental and theoretical research. Indeed an understanding of the current collection process by an electrode is relevant to diverse matters such as Langmuir and Mach-probes calibrations, dusty plasma physics, or spacecraft charging. Recent simulations relying on the ad hoc designed code SCEPTIC have fully addressed the collisionless and unmagnetized problem for a drifting collector idealized as a sphere. SCEPTIC is a 2d/3v hybrid Particle In Cell (PIC) code, in which the ion motion is fully resolved, while the electrons are treated as a Boltzmann distributed fluid [2, 3]. In the present work we tackle the transition between the unmagnetized and the weakly magnetized regime of ion collection by a spherical probe (The mean ion Larmor radius rL > rp) in a collisionless plasma (The ion mean free path Am,fp > rp). When the sphere is at space potential, we demonstrate that the ion current dependence on the background magnetic field B is linear for low B, and provide analytical expressions for this dependence. When the probe potential can not be neglected, the problem shows two distinct scale lengths: A collisionless layer of a few rp close to the probe, followed by a collisional presheath of a few AX,fp. The chosen approach is to resolve the collisionless scale-length with SCEPTIC, while using appropriate outer boundary conditions on the potential and ion distribution function to connect with the unresolved collisional presheath. We present results of our numerical simulations for a wide range of plasma parameters of direct relevance to Langmuir and Mach-probes. / by Leonardo Patacchini. / S.M.

Thesis (Ph. D.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2012. / Cataloged from PDF version of thesis. / Includes bibliographical references (p. 246-253). / For decades, nuclear energy development was based on the expectation that recycling of the fissionable materials in the used fuel from today's light water reactors into advanced (fast) reactors would be implemented as soon as technically feasible in order to extend the nuclear fuel resources. More recently, arguments have been made for deployment of fast reactors in order to reduce the amount of higher actinides, hence the longevity of radioactivity, in the materials destined to a geologic repository. The cost of the fast reactors, together with concerns about the proliferation of the technology of extraction of plutonium from used LWR fuel as well as the large investments in construction of reprocessing facilities have been the basis for arguments to defer the introduction of recycling technologies in many countries including the US. In this thesis, the impacts of alternative reactor technologies on the fuel cycle are assessed. Additionally, metrics to characterize the fuel cycles and systematic approaches to using them to optimize the fuel cycle are presented. The fuel cycle options of the 2010 MIT fuel cycle study are re-examined in light of the expected slower rate of growth in nuclear energy today, using the CAFCA (Code for Advanced Fuel Cycle Analysis). The Once Through Cycle (OTC) is considered as the base-line case, while advanced technologies with fuel recycling characterize the alternative fuel cycle options available in the future. The options include limited recycling in LWRs and full recycling in fast reactors and in high conversion LWRs. Fast reactor technologies studied include both oxide and metal fueled reactors. Additional fuel cycle scenarios presented for the first time in this work assume the deployment of innovative recycling reactor technologies such as the Reduced Moderation Boiling Water Reactors and Uranium-235 initiated Fast Reactors. A sensitivity study focused on system and technology parameters of interest has been conducted to test the robustness of the conclusions presented in the MIT Fuel Cycle Study. These conclusions are found to still hold, even when considering alternative technologies and different sets of simulation assumptions. Additionally, a first of a kind optimization scheme for the nuclear fuel cycle analysis is proposed and the applications of such an optimization are discussed. Optimization metrics of interest for different stakeholders in the fuel cycle (economics, fuel resource utilization, high level waste, transuranics/proliferation management, and environmental impact) are utilized for two different optimization techniques: a linear one and a stochastic one. Stakeholder elicitation provided sets of relative weights for the identified metrics appropriate to each stakeholder group, which were then successfully used to arrive at optimum fuel cycle configurations for recycling technologies. The stochastic optimization tool, based on a genetic algorithm, was used to identify non-inferior solutions according to Pareto's dominance approach to optimization. The main tradeoff for fuel cycle optimization was found to be between economics and most of the other identified metrics. / by Stefano Passerini. / Ph.D.

Thesis: S.M., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 2015. / "June 2015." Cataloged from student-submitted PDF version of thesis. / Includes bibliographical references (pages 70-71). / The offshore floating nuclear power plant (OFNP), is a new power plant design which provides for both increased safety and extra barriers to separate its radioactive material from the public. This design will minimize the probability of a severe accident leading to a release of radioactive material, but as always a release must still be planned for. The offshore siting of an OFNP allows for increased distance to human populations in addition to extra filtering of released radioactive material. This study will look at the potential consequences of a severe accident onboard an OFNP eventually leading to a vent and environment contamination. Three steps of the accident and fallout will be analyzed: 1) Accident and vent composition 2) The transport of radioactive material in the ocean via a plume and ocean diffusion 3) Sedimentation of radioactive cesium on the coast One of the major advantages of an OFNP over a terrestrial plant is that the extra distance and barriers provided by the OFNP will decrease the impact of a nuclear accident. This study will begin to quantify that effect. This is only the first attempt at exploring the effects of a release, and has large conservatisms built into the analysis even in the best estimate case. In the future more detailed work will be done to reach a more accurate solution, particularly for specific siting locations. / by Angelo Briccetti. / S.M.

Thesis (S.B.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2006. / Includes bibliographical references (leaves 73-74). / The objective of this study is to devise a method for utilizing the ORIGEN-S computer code to calculate the activation products generated in in-core experimental assemblies at the MIT Research Reactor (MITR-II). ORIGEN-S is a nuclear depletion and decay analysis code. It accounts for all types of nuclear reactions and eliminates the need for selection of the dominant reactions that will occur in a given experiment, as must be done with the existing activity calculation method. It is expected that the new approach will be easy to use, and will produce radioactivity estimations that are generally more accurate than those produced by the existing method. The ORIGEN-S method has been developed and tested for four experiments that have been or are scheduled to be irradiated in the MITR. These experiments are the Advanced Cladding Irradiation (ACI), High Temperature Irradiation Facility (HTIF), Electric Power Research Institute Electro-Chemical Potential (EPRI ECP) loop, and Annular Fuel Test Rig (AFTR). The method has also been used to perform activation analyses for ten individual elements (plus U-235 and U-238) that are commonly found in MITR in-core experiment (ICE) assemblies. The ORIGEN-S analyses for the ACI, HTIF, and EPRI ECP experiments produced results that were relatively similar to the results produced by previous analyses that utilized the current method of activation estimation. This is because the thermal neutron capture reactions, which are major contributors to the activation of these experiments, are already well accounted for in the existing method. The results of the ORIGEN-S analysis for the AFTR, which contains fissile material, were also very similar to the results of the previous analysis, despite the fact that the previous analysis accounted for changes in flux due to fissile nuclide depletion during irradiation and the current analysis did not. / It is concluded that the activation calculation method developed should be generally adequate for all experiments irradiated in the MITR core. A possible exception involves experiments containing quantities of fissile material larger than the quantities contained in the AFTR, as these experiments could produce significant changes in neutron flux levels that would render this method inadequate. / by Edward M. Helvenston. / S.B.

Thesis (Ph. D.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2009. / Cataloged from PDF version of thesis. / Includes bibliographical references (p. 233-238). / Cable-in-Conduit Conductor is the typical geometry for the conductor employed in superconducting magnets for fusion applications. Once energized, the magnets produce an enormous electromagnetic force and very large transverse loads are applied against the strands. This large force results in a degradation of the performance of the superconducting magnets. In this thesis work transverse load experiments on sub-sized cables, have been designed to study the mechanical and electrical transverse load effects on superconducting cables. Two devices to apply external mechanical loads to a cable have been developed and several different size cables have been tested simulating the International Thermonuclear Experimental Reactor (ITER) Lorentz stress conditions. The first device was designed to use a circular turn sample of a 36-strand cable. Four samples were successfully tested with this device and significant degradations of the critical current due to the external transverse loads have been measured. However, all samples showed unexpectedly large initial degradations that made an analysis of transverse load effects of the samples difficult. The second device was developed for a hairpin configuration. Three different size cables of a single strand, a triplet and a 45-strand cable were systematically tested using this method. This hairpin sample device has successfully operated and provided very reliable experimental data. The experimental results were difficult to explain by existing theories. / (cont.) A new model based on contact mechanics concepts has been developed to determine the number of contacts and the effective contact pressure among the strands in a cable. The model was used to analyze and accurately calculate the displacements of a cable under transverse mechanical load, and it has evaluated the effective contact pressures between strands for the first time. The new model can explain the Lorentz force and contact pressure distribution effect on the critical current degradation of the tested samples. The 3-strand data and their critical current behavior as a function of the effective contact pressure were used to predict the test behavior of a 45-strand cable. It was also used to simulate the critical current degradations of various cables including ITER full size cables. The model has predicted an initial degradation of 20% for an ITER TF cable of 1152 strands at 68 kA operational current caused by the transverse Lorentz load effect only. Parametric studies of the model have indicated that the initial degradation could be reduced by shortening the twist pitch length of the initial stages of a full size cable or by mechanically supporting the last stage bundles of the cable. This thesis work shows for the first time, that the transverse Lorentz load effect, which is inherent in the CICC design, contributes a significant fraction of the degradation of a large Nb3Sn superconducting cable. The model quantifies the degradation and this information could be used in better estimating the appropriate margin requirements in magnet design. / by Luisa Chiesa. / Ph.D.

Thesis (S.B.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2013. / "June 2013." Cataloged from PDF version of thesis. / Includes bibliographical references (pages 32-33). / Given the highly complex nature of neutronics and reactor physics, efficient methods of optimizing are necessary to effectively design the core reloading pattern and operate a nuclear reactor. The current popular methods for optimization are Simulated Annealing and the Genetic Algorithm; this paper explores the potential for a new method called Greedy Exhaustive Dual Binary Swaps (GEDBS). The mandatory trade-off in computation is accuracy for speed; GEDBS is an exhaustive search and tends toward longer runtimes. While GEDBS performed acceptably for the criterion administered in this paper (local peaking and k, on a Boiling Water Reactor (BWR) fuel lattice) the exhaustive nature of GEDBS will inevitably lead to combinatorial explosion for the addition of the potential dozens of factors that commercial application mandates. This issue may be resolved with the addition of metaheuristics to reduce the search space for GEDBS, or by an increasing computation. / by Chad Bums. / 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. 191-197). / Sodium Fast Reactors (SFR) show promise as an effective way to produce clean safe nuclear power while properly managing the fuel cycle. Accurate computer modeling is an important step in the design and eventual licensing of SFRs. The objective of this work was to couple a model for metal fuel performance to a sub-channel analysis code to more precisely predict critical phenomena that could lead to pin failure for steady-state and transient scenarios. The fuel code that was used is the recently developed and benchmarked FEAST-METAL code. The sub-channel analysis code that was selected is COBRA-IV-I. This code was updated with current correlations for sodium for pressure drop, mixing, and heat transfer. The new code, COBRA-IV-I-MIT was then validated with experimental data from the Oak Ridge National Laboratory (ORNL) 19-Pin Bundle, the Toshiba 37-Pin Bundle, and the Westinghouse Advanced Reactors Division (WARD) 61-Pin Bundle. Important topics that were addressed for coupling the codes include the following. The importance of azimuthal effects in the fuel pin: FEAST only evaluates the fuel in two-dimensions, assuming azimuthal symmetry; however, coupling to COBRA produces an azimuthal temperature distribution. The acceptability of assuming a two-dimensional fuel rod with an average temperature was examined. Furthermore, how the fuel pin evolves over time affects the assembly geometry. How well a two-dimensional fuel rod allows for an accurate description of the changing assembly geometry was also considered. Related to this was how the evolution of the assembly geometry affects its thermal hydraulic behavior, which determined the exact form of coupling between the codes. Ultimately one-way coupling was selected with azimuthal temperature averaging around the fuel pin. The codes were coupled using a wrapper, the COBRA And FEAST Executer (CAFE), written in the Python programming language. Data from EBR-II was used to confirm and verify CAFE. It was found that the number of axial nodes used in FEAST can have a large effect on the result. Finally FEAST was used to parametrically study three different pin designs: driver fuel, radial blanket, and tight pitch breed and bum fuel. This study provides data for pin expected life in assembly design. / by Joseph William Fricano. / Ph.D.

Thesis (S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2013. / Cataloged from PDF version of thesis. / Includes bibliographical references (pages 123-133). / Used nuclear fuel dry storage canisters will likely be tasked with holding used nuclear fuel for a period longer than originally intended. Originally designed for 20 years, the storage time will likely approach 100 years. These canisters are fabricated from rolled and welded austenitic stainless steel plate. Most of the storage facilities are located on coastal or brackish water sites with environments containing moisture and chloride ions that can cause stress corrosion cracking (SCC). Residual stresses from the welding process provide the tensile stress for crack initiation and propagation which could eventually compromise canister integrity, allowing the release of radioactive material to the environment. If it is assumed that a tensile stress, predominantly from welding, is constant through the material thickness, this would suggest that failure will be initiation controlled. However, prior studies and practical experience indicate that residual stress varies as a function of depth into a welded material, and that stresses can decrease to zero or even go into compression. This would indicate that at some point, crack propagation could be slowed or even be stopped. In order to predict the time to failure of canister material by stress corrosion cracking, it is therefore necessary to know the actual residual stress distribution through the thickness of canister welds. This thesis investigates dry storage canister designs, canister welds, and contributing factors to residual stress, as well as prior studies of residual stress in welded stainless steel piping and chloride stress corrosion crack propagation rates. From this investigation, an estimate is made for the likely residual stress distribution in a typical canister weld, and the effect of residual stress on canister life prediction is examined. The analysis suggests that residual stress distribution has a tremendous impact on a canister's projected time to failure, and that residual tensile stresses in the heat-affected zone of canister welds could become low enough to result in crack arrest. / by Bradley P. Black. / S.M.