Thesis: S.M., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 2014. / Cataloged from PDF version of thesis. / Includes bibliographical references (pages 140-144). / An individual's perception of radiation, termed as "Radiation Attitudes" in this work, is vital for understanding the stakeholder relationship dynamics for acceptance of controversial nuclear technology projects. Attitudes towards nuclear technology have been found to be different from those towards other technologies perceived as hazardous, such as hydraulic fracturing, genetic engineering or biohazard facilities. Even within the subset of nuclear technology, different applications invoke different reactions. Medical uses of the technology are generally viewed as positive, whereas nuclear power plants and radioactive waste management facilities can sometimes cause fear and anxiety in the minds of some people. This work explains the causes and structure of Radiation Attitudes, and the dynamics of the various factors influencing them. A historical analysis of the narratives concerning nuclear technology was used to identify the complex, social, political, cognitive and technological factors that played a significant role in the formation of Radiation Attitudes. A system dynamics approach was utilized to construct causal loop diagrams depicting the cause-effect relationships and interdependencies between the identified variables. Qualitative interviews were conducted to test the causal relationships hypothesized in the model for Radiation Attitudes. The purpose of the interviews was to understand individual beliefs that result in a particular Radiation Attitude, the bases for these beliefs, and the process of their formation. The interviews enabled verification of the variables and relationships in the model, and the identification of the most significant interdependencies and links. The hypothesized model for Radiation Attitudes correlated well with the information inferred from the interviews, making the first stage of validation a success. / by Aditi Chandra. / 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 (p. 81-83). / The twisting stacked tape cabling (TSTC) method for YBCO superconductors is very attractive for high current density, high magnetic field applications, such as nuclear fusion reactors and high energy physics experiments. Industrial scale assembling methods have been proposed, and cable samples have been tested at 77 K and 4.2 K. A new experimental device has been designed and built to measure critical current of YBCO tapes and TSTC as a function of magnetic field and temperature. The probe allows controlling the temperature between 4.2 K and 80 K within +/-1 K in liquid and gaseous helium ambient, and can be used in a 2 T magnet facility at MIT-PSFC and a 14 T magnet facility at NHMFL-FSU. Its current leads are designed to carry up to 5 kA. The device consists in a 0.9 m long, 25 x 38 mm rectangular vacuum-insulated canister. The superconducting sample and a superconducting current return lead fit inside the canister, in such a way that the Lorentz force and torque produced by the external magnetic field is cancelled. The sample temperature is controlled in a 200 mm long area inside the canister where critical current measurements are performed. Critical current measurements were performed on a single YBCO tape at self-field at temperatures between 20 K and 70 K. The results are similar to data provided by the superconductor's manufacturer. The temperature reached the set point in approximately 10 minutes, and was controlled within +/-1 K. Results of heating power required and difference between set point temperature and measured temperature as functions of set point temperature are presented for two temperature control methods. / by Franco Julio Mangiarotti. / S.M.

Thesis (S.B.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering; and, (S.B.)--Massachusetts Institute of Technology, Dept. of Mathematics, 2011. / "June 2011." Cataloged from PDF version of thesis. / Includes bibliographical references (p. 58-60). / In a system in which metal corrosion is of concern to its long-term structural integrity, crevice corrosion can be a significant cause of damage. Small crevices in a metal exposed to a working fluid (such as a reactor's coolant) may be prone to the development of a localized, aggressive reducing environment. If the metal relies on a passivating layer of oxides for corrosion protection, it may be vulnerable to corrosion attack within the crevice due to a drastically reduced oxygen potential and low pH. Furthermore, in a liquid metal environment, the reducing conditions combined with typically high solubilities of alloy components in the liquid metal can result in severe, localized crevice corrosion that surpasses that which might occur in the aqueous environment of a LWR. In this study, F91 and Fe-12Cr-2Si, two alloys used in previous experiments were exposed to lead-bismuth eutectic maintained at 715*C with a cover gas of pure hydrogen for thirty hours. The conditions were kept extremely reducing, via the initial removal of oxygen and the subsequent maintenance of an environment of pure hydrogen gas, in order to simulate conditions inside a crevice. Following the experiment, the materials were analyzed for corrosion damage via optical microscopy, scanning electron microscopy, and energy-dispersive x-ray spectroscopy. F91 was found to have sustained significant corrosion damage, as expected based on previous experiments, in addition to chromium depletion at the sample surface. Fe-12Cr-2Si was also found to have sustained corrosion damage as a result of lead-bismuth attack. No significant oxide formation or alloying element depletion was observed at the Fe-12Cr-2Si surface. The observed damage in Fe-12Cr-2Si was not entirely expected due to its excellent corrosion resistance in less reducing environments. This raises the concern that crevice corrosion could be an important damage mechanism in applications of the Fe-12Cr-2Si/F91 composite if crevices are present, either due to design flaws or due to cracking during service. / by Sara Elizabeth Ferry. / S.B.

Thesis (Ph. D.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2007. / Includes bibliographical references (leaves 183-188). / Detection of conventional explosives remains a challenge to air security, as indicated by recent reports detailing lapses in security screening and new requirements that mandate screening 100% of checked luggage. Neutron Resonance Radiography (NRR) has been under investigation as a supplement to conventional x-ray systems as a non-invasive, non-destructive means of detecting explosive material in checked luggage. Using fast (1-6 MeV) neutrons produced by an accelerator-based D(d,n)3He reaction and a scintillator-coupled CCD camera, NRR provides both an imaging capability and the ability to determine the chemical composition of materials in baggage or cargo. Theoretical studies and simulations have shown the potential of NRR. This thesis takes the first step towards experimental implementation using a deuterium target for multiple-element discrimination. A new neutron source has been developed to provide the high-flux neutron beam required for NRR while simultaneously minimizing gamma ray production. The gas target incorporates a 4 atm D2 gas chamber, separated from the accelerator beamline with thin, 5 [tm tungsten or 7 [im molybdenum foils supported by a honeycomb lattice structure to increase structural integrity and provide a heat removal pathway. An argon gas cooling system is incorporated to cool the target and thus increase the neutron flux. The gas target has been shown to withstand 3.0 MeV deuteron beam currents in excess of 35 ýLA for extended periods without failure, resulting in a neutron flux of 6.6 x 107 neutrons/sr/LA/s. A neutron imaging system was designed to detect the fast neutrons and produce a digital image of objects for analysis. / (cont.) Two neutron detectors, Eljen plastic scintillator EJ-200 and a ZnS(Ag) scintillating screen were tested for their suitability to NRR. Although ZnS(Ag) has a lower detection efficiency, its resolution, minimal light dispersion, and insensitivity to gamma rays made it the more favorable material. An Apogee Instruments, Inc., Alta U9 CCD camera was used to record the light from the scintillator to create radiographs. The gas target and neutron detection system were used to evaluate the results of experimental work to determine the feasibility of NRR. These experiments ultimately indicated that although NRR has promise, significant challenges regarding neutron flux and image processing must be overcome before the technique can be implemented as an explosives detection system. Suggestions are made for improvements. / by Whitney Lyke Raas. / Ph.D.

Thesis (Ph. D.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2009. / 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. 157-162). / One of the biggest challenges in the physical realization of quantum information processing (QIP) is the precise control of the system. In order to achieve this, we characterize the gates, errors, and noise occurring in experimental setups. In this thesis we develop and further study characterization methods, putting particular emphasis on the scalability problem: O(D4) parameters describe the dynamics of an open quantum system of dimension D, thus O(D4) resources are in principle required to characterize it -- which is a problem in QIP where the desired systems are large (D = 2n for n qubits). We first study the fidelity decay (also called Loschmidt echo) of the system, for many steps under the progressive randomization due to a one-qubit twirl. We show how this quantity encodes useful information about the process begin twirled. We then present a method to measure the magnitude of the multi-body correlations that scales as O(nw), when only up to w-body interactions are expected among the n qubits. We implemented this method in a four-qubit liquid-state Nuclear Magnetic Resonance (NMR) QIP device, demonstrating its potential and feasibility. The experimental work also pointed out the need for robust procedures and the role of implementation errors, while deepening our knowledge of NMR QIP dynamics. We also report on several practical aspects of the experiment, including details on twirls using random rotations and Clifford operators. We furthermore relate this work to recent developments in the community, arriving to a more comprehensive protocol and establishing an intrinsic hierarchy of characterization algorithms. Finally, we study the many-step fidelity decay when using a flawed twirl, thus acknowledging the most realistic scenario where we have a faulty device attempting to characterize itself. Our preliminary work points towards the use of a many-step scheme that promises robust scalable tools to characterize the twirl operators themselves. / by Cecilia Carolina López. / Ph.D.

Thesis (Ph. D.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2006. / "September 2006." In title on t.p.,double-underscored letters "t" and "h" appear as subscript. / Includes bibliographical references (p. 229-233). / The gas cooled fast reactor (GFR) has received new attention as one of the basic concepts selected by the Generation-IV International Forum (GIF) for further investigation. Currently, the reference GFR is a helium-cooled direct cycle plant with core outlet temperatures in the 850"C to 10000C range. Pursued in the interest of high cycle efficiency and the provision of process heat for hydrogen production by thermochemical water cracking, these high temperatures present materials challenges which may prove difficult to overcome in the near future. By taking advantage of the low compressibility of CO2 near its critical point, the supercritical CO2 (S-CO2) recompression cycle can achieve an efficiency of 48% with a relatively low core outlet temperature of 650'C. The 4-loop 2400 MWth direct S-CO2 cooled fast reactor under investigation at MIT is thus a lower-temperature alternative to the mainstream helium cooled GFR design. A steady state core design was developed which utilizes an innovative, high fuel volume fraction, vented Tube-In-Duct (TID) fuel assembly. Through an extensive series of iterative calculations, RELAP5-3D was then used to evaluate the natural circulation performance of an active/passive hybrid Shutdown/Emergency Cooling System (SCS/ECS). Routes were identified by which significant post-LOCA core bypass could occur and degrade the decay heat removal performance. Moderately-sized blowers were shown to be capable of overcoming even extreme core bypass routes. An active SCS/ECS was thus adopted for the reference design. / (cont.) The loss of external load (LOEL) event is analyzed and a bypass valve scheme is recommended which prevents shaft overspeed and excessive core coolant mass flow rate. A large dry pressurized water reactor (PWR) containment building having a free volume of 70,000 m3 and a peak design pressure of 6 bar is selected for this design based on a 100 in2 cold duct break. During this same loss of coolant accident (LOCA), the depressurization time is shown to be in excess of 10 minutes. No action need be taken by the SCS/ECS blowers before this time in order to prevent core damage. After this time, a total blower power less than 90 kW is sufficient to cool the core out to 10,000 seconds. A loss of flow (LOF) transient in which a PCS loop is instantaneously isolated and no mitigating action is taken (i.e. no reactor scram) is also shown not to cause core damage. It is concluded that a large S-CO2 cooled GFR coupled to a supercritical Brayton power conversion system can withstand the thermal hydraulic challenges posed by the usual menu of severe accident scenarios. / by Michael A. Pope. / Ph.D.

Thesis (S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2011. / Cataloged from PDF version of thesis. / Includes bibliographical references (p. 146-149). / High conversion water reactors (HCWR), such as the Resource-renewable Boiling Water Reactor (RBWR), are being designed with axial heterogeneity of alternating fissile and blanket zones to achieve a conversion ratio of greater than one and assure negative void coefficient of reactivity. This study assesses the generation of few-group macroscopic cross sections for neutron diffusion theory analyses of this type of reactor, in order to enable three-dimensional transient simulations. The goal is to minimize the number of energy groups in these simulations to reduce computational effort. A two-dimensional cross section generation methodology using the Monte Carlo code Serpent, similar to the traditional deterministic homogenization methodology, was used to analyze a single RBWR assembly. Results from two energy group and twelve energy group diffusion analyses showed an error in multiplication factor over 1000 pcm with errors in reaction rates between 10 and 60%. Therefore, the traditional approach is not sufficiently accurate. Instead, a three-dimensional homogenization methodology using Serpent was developed to account for neighboring zones in the homogenization process. A Python wrapper, SerpentXS, was developed to perform branch case calculations with Serpent to parametrize few-group parameters as a function of reactor operating conditions and to create a database for interpolation with the nodal diffusion theory code, PARCS. Diffusion analyses using this methodology also showed an error in multiplication factor over 1000 pcm. The three-dimensional homogenization capability in Serpent allowed for the introduction of axial discontinuity factors in the diffusion theory analysis, needed to preserve Monte Carlo reaction rates and global multiplication factor. A one-dimensional finite-difference multigroup diffusion theory code, developed in MATLAB, was written to investigate the use of axial discontinuity factors for a single RBWR assembly. The application of discontinuity factors on either side of each axial interface preserved multiplication factor and reaction rate estimates between transport theory and diffusion theory analyses to within statistical uncertainty. Use of this three-dimensional assembly homogenization approach in generating few-group macroscopic cross sections and axial discontinuity factors as a function of operating conditions will help further research in transient diffusion theory simulations of axially heterogeneous reactors. / by Bryan R. Herman. / S.M.

Thesis (S.M. and S.B.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2006. / Includes bibliographical references (p. 179-183). / To increase the power density and maximum allowable fuel burnup in light water reactors, new fuel rod designs are investigated. Such fuel is desirable for improving the economic performance light water reactors loaded with transuranic-bearing fuel for transmutation, as well as those using UO2 fuel. A proposal for using silicon carbide duplex as fuel cladding is investigated. The cladding consists of a monolithic inner layer surrounded by a tightly wound fiber-matrix composite. The monolith layer retains the volatile fission products while the composite adds strength. The FRAPCON steady-state thermo-mechanical fuel rod modeling code is used to examine the performance of SiC cladding at high fuel burnup and high power density. Empirical models are developed to describe the physical properties of the composite as a function of operating temperature and neutron fluence. A comparison of the behavior of the SiC cladding to the conventional Zircaloy cladding demonstrates that the SiC has superior resistance to creep and mechanical degradation due to radiation or oxidation. However, the lower thermal conductivity of the SiC is a major issue, which results in significantly increased peak fuel temperatures. Mixed U02-PuO2 fuel is also examined in place of traditional UO2 pellets, since this may better resemble transmutation fuels of the future. It is found that the use of plutonium-bearing mixed-oxide fuels further exacerbates the high fuel temperatures. The silicon carbide cladding is predicted to have more favorable performance when used for internally- and externally-cooled annular fuel rods developed at MIT. Both sintered annular pellets and VIPAC granular fuel are examined. / (cont.) Because of the fuel geometry, the average fuel temperature is significantly lower, and the stiffness of the SiC cladding helps to maintain the geometry of the annulus during extended irradiation. Experimental projects have been undertaken to study the performance of both the annular fuel rods and silicon carbide duplex cladding. A post-irradiation examination of prototype annular fuel rods with VIPAC fuel, irradiated in the MIT reactor, has been designed and executed. Through this non-destructive examination, the disposition of the fuel grains is examined, and fuel burnup and fission gas release is estimated. These experimental results correlate well with computer calculations. A new irradiation facility was also planned and constructed that consists of a closed loop, operated at pressurized water reactor pressure, temperature, and chemistry conditions. This facility contains silicon carbide duplex cladding samples of various constructions, and it will be irradiated in the core of the MIT reactor for several months. / by David Michael Carpenter. / S.M.and S.B.

Thesis (S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2011. / Cataloged from PDF version of thesis. / Includes bibliographical references (p. 215-218). / The most prevalent criticism of fission-fusion hybrids is simply that they are too exotic - that they would exacerbate the challenges of both fission and fusion. This is not really true. Intriguingly, hybrids could actually be more viable than stand-alone fusion reactors while mitigating many challenges of fission. This work develops a conceptual design for a fission-fusion hybrid reactor in steady-state L-mode tokamak configuration with a subcritical natural or depleted uranium pebble bed blanket. A liquid lithiumlead alloy breeds enough tritium to replenish that consumed by the D-T fusion reaction. Subcritical operation could obviate the most challenging fuel cycle aspects of fission. The fission blanket augments the fusion power such that the fusion core itself need not have a high power gain, thus allowing for fully non-inductive (steady-state) low confinement mode (L-mode) operation at relatively small physical dimensions. A neutron transport Monte Carlo code models the natural uranium fission blanket. Maximizing the fission power while breeding sufficient tritium allows for the selection of an optimal set of blanket parameters, which yields a maximum prudent fission power gain of 7.7. A 0-D tokamak model suffices to analyze approximate tokamak operating conditions. If the definition of a "reactor" is a device with a total power gain of 40, then this fission blanket would allow the fusion component of a hybrid reactor with the same dimensions as ITER to operate in steady-state L-mode very comfortably with a fusion power gain of 6.7 and a thermal fusion power of 2.1 GW. Taking this further can determine the approximate minimum scale for a steady-state L-mode tokamak hybrid reactor, which is a major radius of 5.2 in and an aspect ratio of 2.8. This minimum scale device operates barely within the steady-state L-mode realm with a thermal fusion power of 1.7 GW. This hybrid, with its very fast neutron spectrum, could be superior to pure fission reactors in terms of breeding fissile fuel and transmuting deleterious fission products. It could operate either as a breeder, producing fuel for pure fission reactors from natural or depleted uranium, or as a deep burner, fissioning heavy metal and transmuting waste with a cycle time of decades. Despite a plethora of potential functions, its primary mission is deemed to be that of a deep burner producing baseload commercial power with a once-through fuel cycle. Although hybrids are often purported a priori to pose an elevated proliferation risk, this reactor breeds plutonium that could actually be more proliferation-resistant than that bred by fast reactors. Furthermore, a novel method (the "variable fixed source method") can maintain constant total hybrid power output as burnup proceeds by varying the neutron source strength. As for engineering feasibility, basic thermal hydraulic analysis demonstrates that pressurized helium could cool the pebble bed fission blanket with a flow rate below 10 m/s. The Brayton cycle thermal efficiency is 41%. This device is dubbed the Steady-State L-Mode Non-Enriched Uranium Tokamak Hybrid (SLEUTH). The purpose of this work is not any sort of elaborate design, but rather the exploration of an idea coupled with corroborating numerical analysis. At this point in the hybrid debate, viable conceptual designs are persuasive while intricate build-ready designs are superfluous. This work conceives such a conceptual design, demonstrates its viability, and will perhaps, incidentally, spur a profusion of pro-fusion sentiment! / by Mark Wilbert Reed. / S.M.

Thesis (S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2012. / Cataloged from PDF version of thesis. / Includes bibliographical references (p. 150-157). / Previous studies conducted at MIT showed that the power performance of an inverted pressurized water reactor (IPWR) conceptual design, i.e. the coolant and moderator are inverted such that the fuel is the continuous medium and the moderator flows through coolant channels, has potential to outperform a traditional pressurized water reactor (PWR). Similar to the traditional PWR, the IPWR design involves a tradeoff between core pressure drop and the minimum departure from nucleate boiling ratio (MDNBR). In order to increase the power density of the IPWR, Ferroni [231 examined the possibility of inserting multiple short-length twisted-tapes (MSLTTs) in the cooling channels. For a fixed coolant mass flow rate, the swirling flow produced by the MSLTTs allows the IPWR to have a higher operating heat flux while maintaining the design criteria of MDNBR as compared to either the traditional PWR or IPWR without swirl promoters. However, the addition of each twisted-tape increases the core pressure drop which limits the coolant flow rate due to pumping power limitations of existing reactor coolant pumps (RCPs). In order to better characterize the critical heat flux (CHF) enhancement caused by the addition of MSLTTs, this study performed a critical analysis of existing CHF correlations and models. Initially a phenomenological model was sought to describe the mechanisms of CHF for tubes containing MSLTTs; however, the full-length twisted-tape (FLTT) model that was selected for modification was found to have terms that could not be reconciled for the transition from fully developed swirl to decaying swirl. The existing CHF correlations for swirling flow were also found to be unsatisfactory. Therefore, the insights gained through working with the phenomenological model were used to develop a new empirical correlation to describe the departure from nucleate boiling (DNB) using existing swirling flow DNB data as well as an existing swirl decay model. In order to allow for more flexibility in the placement of the MSLThs, an existing FLTT pressure drop correlation was modified to account for the form pressure drop at the entrance to each twisted-tape insert as well as the friction pressure drop in the decaying swirl region downstream from the exit of each MSLTT. A sensitivity analysis of the new pressure drop correlation was also performed to determine if the complete methodology could be simplified. Design insights were presented that help to narrow the design space for the IPWR. These steps should be followed in order to find the maximum power density possible by the IPWR design. Finally, the existing swirl flow CHF data and correlations are presented in the appendices of this thesis. / by Tyrell Wayne Arment. / S.M.