Thesis (S.B.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2006. / Includes bibliographical references (p. 53-57). / India is in the middle of the biggest expansion of nuclear power in its history, adding 20 GWe in the next 14 years in the form of pressure water reactors and fast breeder reactors. At the same time, the United States is overturning decades of policy in order to resume the export of nuclear materials to India, opening up the possibility of private investors in the Indian nuclear industry for the first time. This is a period of progress and turmoil in India's nuclear power program. This thesis seeks to describe and analyze India's nuclear prospects and to qualitatively assess the system's strengths and weaknesses. Using the inception of the country's nuclear power program as a starting point, this thesis will trace India's nuclear lineage to the present. In the process, it will evaluate what makes the Indian program unique, and why it may not be ideal for India that the United States is finally renewing its offers of a cooperative nuclear alliance. / by Caitlin Lenore Murray. / S.B.

Thesis (Ph. D.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2011. / Cataloged from PDF version of thesis. / Includes bibliographical references (p. 262-269). / This study assesses the neutronic, thermal-hydraulic, and fuel performance aspects of using nitride fuel in place of oxides in Pu-based high conversion light water reactor designs. Using the higher density nitride fuel hardens the neutron energy spectrum and results in higher breeding ratios. The state-of-the-art high conversion light water reactor, the Resource-renewable Boiling Water Reactor (RBWR), served as the template core upon which comparative studies between nitride and oxide fuels were performed. A 1/3 core reactor physics model was developed for the RBWR using the stochastic transport code MCNP. The code was coupled with a lumped channel thermalhydraulics 5-channel model for steady-state analyses. The depletion code MCODE, which links MCNP with ORIGEN, was used for all burnup calculations. Select physics parameters were calculated and with the exception of the void coefficients, agreed with reported data. The void coefficients of the coupled core were calculated to be slightly positive using two different methods (10% power increase and 5% flow reduction). The standard RBWR assembly designs, which use tight lattice hexagonal fuel rod arrays, with oxide fuel were then replaced with various nitride fuel assembly designs to determine the potential increase in breeding ratio, the potential to breed with pressurized water, and the potential to improve the critical power ratio with a wider pin pitch. Without changing the assembly geometry or discharge burnup, using nitride fuel resulted in a breeding ratio of 1.14. Using single-phase liquid water, the nitride fuel RBWR assembly resulted in a conversion ratio of 1.00. Another nitride fuel assembly design with boiling water maintained a 1.04 breeding ratio while increasing the pitch-todiameter ratio from 1.13 to 1.20. This modification increased the hot assembly critical power ratio from 1.22 to 1.36, as calculated using the Liu-2007 correlation. A high-porosity nitride fuel is recommended for high burnup conditions, to accommodate the nitride fuel's higher swelling and less favorable mechanical properties compared to the oxide fuel. The high porosity allows additional volume for pressure-induced densification, alleviating swelling and subsequent cladding strain. To predict the performance of high-porosity nitride fuel, fission gas and fuel behavior mechanistic models were developed for high burnup and low-temperature conditions. These models were validated with reported irradiation data and implemented, along with fuel material properties, into the steady-state fuel behavior code FRAPCON-EP. Under simulated RBWR conditions, a fuel density no more than 85% of theoretical density is recommended to maintain satisfactory fuel performance. / by Bo Feng. / Ph.D.

Thesis (Ph. D.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2009. / Cataloged from PDF version of thesis. / Includes bibliographical references (p. 189-197). / It is commonly accepted that ExB velocity shear is responsible for the suppression of edge turbulence, which reduces the losses of both energy and particles across magnetic field lines and results in the formation of edge transport barriers and high-confinement mode (H-mode) in tokamak plasmas. However, the self consistent evolution of the radial electric field profile (Er), pedestal shape and improvement in plasma confinement is not well understood. A better understanding of pedestal physics and the interplay between Er, turbulence suppression and pedestal formation should enable better control of edge transport and improve core confinement. A new, high-resolution, charge exchange recombination spectroscopy (CXRS) diagnostic has been installed on Alcator C-Mod to provide measurements of the B5+ population in the pedestal region. This diagnostic is capable of measuring the boron temperature, density, and poloidal and toroidal velocity with 3mm radial resolution and 5ms temporal resolution. These profiles, coupled with knowledge of the toroidal and poloidal magnetic fields, enable the determination of the edge radial electric field through the radial force balance equation. The new CXRS diagnostic has provided the first spatially resolved calculations of the radial electric field in the C-Mod edge and has made possible significant contributions to the study of pedestal physics. Detailed measurements of the boron population have been made in a variety of plasma regimes. The measured rotation profiles connect the SOL and core measurements and are consistent with both. The CXRS boron temperature profiles are observed to agree well with the Thomson Scattering electron temperature profiles in bothl shape and magnitude over a wide range of collisionalities. In H-mode plasmas both the boron temperature and density profiles form clear pedestals, similar to what is observed in the electron channel. The edge toroidal rotation increases in the concurrent direction at the onset of H-mode confinement and the poloidal rotation in the pedestal region increases in the electron diamagnetic direction forming a narrow / (cont.) peak (3-4mm) just inside of the LCFS. In Ohmic L-mode plasmas Er is positive near the last closed flux surface (LCFS) and becomes more negative with distance into the plasma. In H-mode plasmas E, is positive in the core, but forms a deep negative well, relative to its L-mode values, just inside of the LCFS. These results are qualitatively consistent with the observations made on other machines. However, the C-Mod H-mode Er wells are unprecedeited in depth (up to 300kV/m) and the narrow E, well widths (5mm), as compareJ to results from other tokamaks, suggest a scaling with machine size. The measured Er well widths have been compared to theoretical scalings for the edge pedestal and no significant correlation was observed with any of the predictions. In fact, very little variation of the E, well width is observed in general. Howc:ver, the depth of the E, well, or alternatively the magnitude of the E, shear (constant width), shows a strong correlation with improved plasma energy confinement. It also correlates well with the edge electron temperature and pressure pedestal heights (and gradients). It is not, however, very sensitive to variation in the edge electron density pedestal height. These results are an indication that the energy and particle transport have different relationships to Er, with energy transport more directly linked. The radial electric field results from ELM-free H-mode and I-mode plasmas support this interpretation. / by Rachael Marie McDermott. / Ph.D.

Thesis: Sc. D., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 2018. / Cataloged from PDF version of thesis. "February 2018." / Includes bibliographical references (pages 203-214). / Using high-z wall materials switches the fusion challenge from heat load handling to removing impurities. I propose the first method of measuring the radial impurity flux from currently available diagnostics. It provides a means of solving the impurity accumulation problem while providing free fueling for optimum tokamak performance. High confinement mode operation was discovered 35 years ago to almost quadruple fusion power, and later explained by turbulence reduction by sheared flows. Less than a decade ago, improved mode operation was discovered to have the same desirable property, while removing impurities and providing fueling. Thanks to the impurity radial particle flux measuring technique developed, I explain the outward radial impurity flux without invoking a (sometimes undetected) turbulent mode. This theory is supported by the observed E x B flow shear, which also explains the desired energy confinement via turbulence reduction. Stronger impurity density in-out poloidal asymmetries than predicted by the most comprehensive neoclassical models have been measured in several tokamaks around the world during the last decade, calling into question the reduction of turbulence by sheared radial electric fields in H-mode tokamak pedestals. However, these pioneering theories neglect the impurity diamagnetic drift, or fail to retain it self-consistently as proven in this thesis, while recent measurements indicate that it can be of the same order as the ExB drift. I have developed the first self-consistent theoretical model retaining the impurity diamagnetic flow and the two-dimensional features it implies due to its associated non-negligible radial flow divergence. It successfully explains collisionally the experimental impurity density, temperature and radial electric field in-out asymmetries; thus making them consistent with H-mode pedestal turbulence reduction. / by Silvia Espinosa Gútiez. / Sc. D.

Thesis (S.M. and S.B.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2008. / Includes bibliographical references (leaves 149-150). / Recently, MIT's Center for Advanced Nuclear Energy Systems developed a new high burnup annular fuel that features both internal and external cooling. Implementation of this fuel design in current pressurized water reactors (PWRs) will allow power uprates up to 50% while maintaining or improving its existing thermal and safety margins. Each annular fuel assembly is arranged in a 13x13 array but has the same side dimensions as a 17x17 solid fuel assembly. Even at much higher power densities, the peak fuel temperatures are substantially lower and the MDNBR is comparable to that of solid fuel at 100% power. The main motivation for utilizing this fuel is the lower capital construction cost per kilowatt of electrical production compared to new reactors using solid fuel. To elaborate on the previous work, three remaining issues were addressed: the shutdown margin deficiency at 50% uprated power, effect of inner channel flow restrictions due to crud buildup and obstructions, and the economic impact of a fleet of reactors using high burnup annular fuel. All of the work was done using computer codes specializing in core neutronics, thermal hydraulics, and fuel cycle analysis. The reduced shutdown margin was found to be caused mainly by a reduction in control material volume coupled with a higher power density. This issue was resolved by changing the control material from traditional Ag-In-Cd to 25 wt% B-10 enriched B₄C. Increasing the control rod surface area was also investigated as a possible solution but it was revealed that any departure from the cylindrical shape would lead to a reduction in control volume which resulted in decreased rod worth. Simultaneous oxide growth and crud buildup on the inner cladding of the annular fuel was simulated in a whole core thermal-hydraulics model to determine the maximum thickness that the annular fuel can tolerate while maintaining an MDNBR greater than 1.3 under transient overpower conditions. / (cont.) Under very conservative conditions, the maximum tolerable thickness was calculated to be a uniform 50 [mu]m layer of combined oxide and corrosion on the inner and outer cladding surfaces of the hot rod. Under full power conditions, the fuel was found to be able to tolerate a 35-40% blockage of the hot rod's inner channel. However, the prospect of plugging can be regarded as very hypothetical since debris filters in PWRs have a mesh spacing smaller than the fuel's inner channel diameter. The fuel cycle analysis code CAFCA SD was modified to include the capability to model the effect of LWRs using high burnup annular fuel on the US fuel cycle. At a given date, annular fuel can be introduced to the once-through fuel cycle via Generation II reactor uprates or construction of Generation Ill reactors. Results showed that constructing new 1.5 GWe reactors using annular fuel resulted in the greatest reduction in the cost of electricity due to its low capital construction cost. Total spent fuel was also reduced due to the reduced amount of reactors required to fulfill power demand. However, compared to traditional 4.5 wt% enriched solid fuel, using the higher enriched annular fuel for an entire fleet of LWRs would require a greater uranium mining rate. Overall, all three of these studies alleviate some concerns about annular fuel and serve to boost its attractiveness and feasibility for use on an industrial scale. / by Bo Feng. / S.M.and S.B.

Thesis (S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2007. / Includes bibliographical references (p. 76). / The glaring need to develop methods for detecting and interdicting illicit nuclear trafficking has resulted in the exploration of various methods for active neutron interrogation, specifically for the presence of special nuclear material (SNM) in cargo containers. The proposed system aims to defeat the ability of terrorists to import SNM into the United States via maritime shipping, thus greatly reducing the possibility of a successful nuclear terrorist attack. The proposed system uses 60-100 keV neutrons, produced by the 7Li(p,n)7Be reaction in a linear accelerator and kinematically beamed into various targets. In the event that fissile material is present, highly energetic neutrons will be emitted from the fissioning of a nucleus and some of these neutrons will eventually radiate from the container. Inevitably, high energy photons will also radiate from the target due to the interactions of neutrons and host materials. Utilizing a neutron detection system that is able to discriminate low energy neutrons, high energy gamma rays and the high energy neutrons from fission enables the detection of fissile material in various containers. An increase in discriminated high energy neutron events during active neutron interrogation selectively indicates the presence of SNM, since neutron energies on the order of 1 MeV are required for the SNM-equivalent fissioning potential when incident upon U238 and other high-Z nuclei. Furthermore, neutrons with less than approximately 100 keV do not undergo nuclear processes such as (n,2n) and (n,n'), but rather lose their energy through kinematic collisions. / (cont.)Results obtained validate this proof-of-concept, in that observed high energy neutron events increase significantly in the presence of gram-quantities of SNM. Further, attempts made to shield the SNM from active interrogation do not defeat the proposed system's ability to identify the presence of SNM. With a fully-functional proof-of-concept, further work towards developing a complete and deployable prototype active neutron interrogation system will serve to augment the ability of the United States to detect, deter and interdict illicit nuclear trafficking. / by Eduardo A. Padilla. / S.M.

Thesis (S.M. and S.B.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2009. / Cataloged from PDF version of thesis. / Includes bibliographical references (p. 59). / The International Atomic Energy Agency requires nuclear facilities around the world to abide by heavily enforced safeguards to prevent proliferation. Nuclear fuel reprocessing facilities are designed to be proliferation-resistant and to use surveillance systems. While experience with small-scale reprocessing facilities has allowed for well understood safeguards, large-scale reprocessing facilities pose a new difficulty because of the larger error margins involved with the large volumes of spent fuel that is being processed. First, a hypothetical spent nuclear fuel reprocessing facility is described along with proliferation resistance methods typically used in actual facilities. This model establishes a foundation for studying diversion scenarios using a success tree method. / by Lindsay Lorraine Calderón. / S.M.and S.B.

Thesis (S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2010. / 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 (p. 81-82). / A unique methodology for creating a neutron source model was developed for deuterons and protons incident on solid phase beryllium and lithium targets. This model was then validated against experimental results already available in literature. The model was found to adequately characterize the deuteron reactions, subject to the end user's desired tolerances, but only marginally so for the proton reactions. A nonstandard, yet practical, application of such a neutron source model was demonstrated by conducting activation analysis for materials reasonably likely to be components in an accelerator based neutron source. This analysis consisted of inducing activation by exposure in the MITR-II 5 MW research reactor, and then adjusting the results based on the model's calculated neutron response. The analysis thus illustrated the utility of the source model in characterizing the system, enabling component design and evaluation before fabrication. / by Peter H. Chapman. / S.M.

Thesis: S.M. and S.B., 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 63-64). / An operating nuclear power reactor is a complex system that is sensitive to many material parameters including densities, temperatures, and compositions. There is great interest in solving the neutron transport with Monte Carlo methods due to their extremely high fidelity, but Monte Carlo methods are too slow to run in an iterative brute-force search of the reactor parameter space. This thesis discusses the derivation, implementation, and applications of differential tallying -- a method which can be used to mitigate the computational cost of mapping out a reactor parameter space with Monte Carlo. With differential tallies, each calculation provides derivatives of tallied quantities like reactivity and fission reaction rates with respect to material density, temperature, etc. These derivatives directly provide reactivity coefficients and they can also be used to extrapolate and predict small changes in reactor parameters. Notably, a novel method is presented which uses the windowed multipole cross section representation to compute temperature derivatives due to the resolved resonance Doppler broadening effect. To demonstrate the utility of differential tallies, this thesis presents example computations of moderator density and fuel Doppler feedback coefficients in pressurized water reactor pincells. With differential tallies, the moderator and fuel Doppler coefficients can be computed 40% and 50x faster, respectively, than by brute-force methods. A calculation of pin-by-pin Doppler coefficients in an assembly is also presented in order to demonstrate that differential tallies are even more efficient for assembly calculations. / by Sterling Harper. / S.M. and S.B.

Thesis (S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2006. / Includes bibliographical references (leaves 197-199). / Superconducting magnets will play a central role for the success of the International Thermonuclear Experimental Reactor (ITER). ITER is a current driven plasma experiment that could set a milestone towards the demonstration of fusion as a source of energy in the future. Cable-in-Conduit is the typical geometry for the conductor employed in superconducting magnets for fusion application. The cable is composed of over 800 strands. Once energized, the magnets produce an enormous electromagnetic force defined by the product of the current and the magnetic field. The strands move under the effect of this force, and the force accumulates against one side of the conduit thereby pressing transversally against the strands. The experiment proposed here has the goal of assessing the functionality of the apparatus designed to study the effect of transverse load on a cable composed of 36 superconducting strands (with a 3x3x4 pattern) by mechanically simulating the ITER Lorentz stress condition. The apparatus was assembled at MIT and preliminary tests at 77 K and room temperature were made to improve the design prior to carrying out the actual experiments. These were done at the National High Magnetic Field Laboratory (NHMFL) located in Florida. Ideally, the transverse conditions simulating the ITER conditions should be created by Lorentz forces due to current and magnetic field. Unfortunately to create such a high level of stress, currents higher than the power supply capability at NHMFL (10 kA) would be required. This is the driving reason to have an apparatus simulating the same stress condition mechanically. / The first test was conducted in October 2005. It was possible to test the structure and its range of operation. Critical current measurements were made as a function of different fields. However during the first measurement, under the loading conditions, the sample was irreversibly damaged and no other measurements were possible. The successful test of the structural behavior of the apparatus motivated a second test carried out in January 2006. With the improvements made between the two experiments, it was possible to successfully measure the degradation of the cable as a function of the transverse pressure applied, measuring degradation as high as 50% with a transverse load of 100 MPa. The ultimate goal of these studies is to characterize the critical current behavior as a function of transverse load in order to predict the response of a full sized Cable-in-Conduit. The work in this thesis was used to explore a setup for measurements and measurement technique. A set of empirical equations describing the behavior of full size cables is needed and should be addressed with a new project that extends the work done so far. / by Luisa Chiesa. / S.M.