Thesis: S.M., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 2015. / Cataloged from PDF version of thesis. / Includes bibliographical references (pages 91-92). / In this thesis, I present a new diagnostic technique for interrogating boiling heat transfer phenomena. The technique, called Multi-Spectral Infrared Thermography or MiST, builds on previous diagnostic techniques for measuring the 2-D wall temperature distribution or the 2-D phase distribution of the fluid above the surface. These methods make use of infrared thermography, a well developed practice involving the use of a high-speed infrared camera to collect visual data. By analyzing the data with both qualitative and quantitative tools, insights into boiling heat transfer mechanisms can be gained. In addition to the MiST technique, a refined infrared camera calibration model is presented for accurately determining the wall temperature. MiST is a new technique that allows for the simultaneous measurement of both the temperature distribution and the phase distribution. This is in sharp contrast to previous techniques which have only allowed the measurement of one or the other. MiST uses a highly engineered, semi-transparent, thin-film heater to enable the simultaneous measurement of the two properties. The heater separates the two signals, one from the temperature and one from the phase, by taking advantage of two regions of the electromagnetic spectrum. By spectrally separating the two signals, no limitation in resolution or field of view is made. The refined camera calibration model presented builds on previous work, which quantified the radiation captured by the camera and used a coupled radiation and conduction model to back out the complete axial temperature distribution within the heater. The new model refines the older version by taking into account spectrally varying optical properties within the heater. The spectral data is easily acquired with a Fourier transform infrared spectrometer, and fed into the radiation model for enhanced accuracy. The development of MiST presents new opportunities in boiling heat transfer for insight into a complex phenomena. The use of MiST in boiling and condensation experiments will lead to the development of new heat transfer models, and can provide high-resolution data for computational fluid dynamics models. MiST presents the logical progression forward in boiling diagnostic tools as it provides enhanced data acquisition opportunities when compared to it's legacy versions. / by Andrew Jonathan Richenderfer. / 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. 117-118). / Standoff detection of improvised explosive devices (IEDs) is a continuing problem for the U.S. military. Current X-ray detection systems cannot detect explosives at distances above a few meters and with a source-detector system moving in relation to the target. The aim of this study is to determine the feasibility of a large-area, Coded-Aperture Imaging (CAI) system using X-Ray backscatter as the source of radiation. A moving source-detector system required development of a new reconstruction technique, dynamic reconstruction (DR), which continually back-projects detected events on an event-by-event basis. This research imaged multiple low-Z (polyethylene and water-filled), area targets with backscattered X-rays using standard medical imaging equipment, coded aperture masks with ideal bi-level autocorrelation properties, and dynamic reconstruction (DR). Lower fill factor apertures were the primary metric investigated because contrast was shown to be inversely related to the mask's percentage of open area. This study experimentally determined the optimal mask fill factor, gamma camera imaging protocols, and experimental geometry by examining the resulting effects on image quality. Reconstructed images were analyzed for Contrast-to-noise ratio (CNR), Signal-to-noise Ratio (SNR), resolution, sharpness, the uniformity of the background (artifacts). In addition to changing the fill factor, additional methods of improving the contrast included changing the experimental geometry, reducing the X-ray tube filtration, and widening the X-ray source's cone beam (FOV). / (cont.) 14 studies were performed that found 25% fill factor mask reconstructions had the highest average CNR (14.7), compared to 50% and 12.5% fill factor (CNRs 8.50 and 6.9, respectively) with a system resolution of 25 mm at the target. Thus, this study's techniques confirmed that large-area, low fill factor coded apertures could successfully be used, in conjunction with dynamic reconstruction, to image complex, extended scenes at 5 meters with capabilities of up to 50 meters or more. / by Jayna T. Bell. / S.M.

Thesis (Sc. D.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2013. / Cataloged from PDF version of thesis. / Includes bibliographical references (pages 171-174). / The effect of aging on nuclear power plant performance has come under increased scrutiny in recent years. The approaches used to make an assessment of this effect strongly influence the economics of nuclear power plant operation. This work aims (a) to determine how the risk-informed approach to performance assessment can be specifically improved, (b) to formulate and demonstrate a method for risk-informed nuclear power plant component reliability analysis, and (c) to assess the applicability of these improvements to the nuclear regulatory environments of the United States and France, the countries having the largest national nuclear power programs. In order to address the first objective, a case study of the U.S. and French regulatory treatment of the pressurized thermal shock phenomenon was performed. A new methodology that can be applied to the construction of risk-informed analyses was developed in response to the findings of the case study. The methodology specifically aims to improve upon current risk-informed methods, which implement a hybrid of best-estimate and conservative modeling, by increasing the use of risk-based modeling justification in order to allow for the development of a more logically consistent analysis. The second objective is addressed through the development of a new reliability model describing the effects of transient-induced degradation. The reliability model is applied through the demonstration of a risk-informed method for the calculation of fatigue degradation. The data requirements for the future successful implementation of this risk-informed method are presented. Examining the applicability of the risk-informed approach to U.S. and French nuclear power plant regulation showed that currently the U.S. is a more progressive implementer of risk-informed practices. This examination revealed, however, than an alternative approach known as the best-estimate plus uncertainty (BEPU) method could be considered for application in regulatory environments, such as in France, that are less accepting of the risk-informed approach. In order to understand how the BEPU method could be improved, the U.S. approach to BEPU was analyzed and recommendations for its future application are presented. / by Brittany L. Guyer. / Sc.D.

Thesis: S.B., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 2014. / Cataloged from PDF version of thesis. / Includes bibliographical references (pages 30-31). / The ultimate storage destination for spent nuclear fuel in the United States is currently undecided. Spent fuel will be stored indefinitely in dry cask storage systems typically located on-site at the reactor or at a dedicated independent spent fuel storage installation (ISFSI). Since these canisters were not originally designed or qualified for indefinite storage, there is a need to quantify the length of time they will be viable for storing spent fuel. Stress corrosion cracking (SCC) is a concern in these canisters if they are exposed to an aqueous, chloride-containing film. Canisters are fabricated using a concrete overpacking, along with austenitic stainless steel on the inside which is welded together. One factor that would significantly impact SCC behavior inside these canister welds, if the proper conditions developed such that SCC occurred, is the tensile residual stress profile. As the highest residual stresses are present in the welds and their heat-affected zones (HAZ), it would be useful to investigate their influence by predicting the residual stress profile in the container. These data will support further research into the life expectancy of these canisters and the possible ways in which they might fail due to SCC. Residual stress data for nuclear waste canisters are scarce. Without experimental measurements, initial insight must be attained through computational analysis using finite-element analysis (FEA) packages such as COMSOL. Using a representative 304L/308L weld plate as a model in COMSOL, predicted residual stress shows some agreement with expected trends: high tensile stresses in the weld/ HAZ regions and compressive stresses in the surrounding material. Hardness tests show trends similar to the hardening profiles that were created after the weld simulation. Additionally, the thermal model may offer insight in predicting the HAZ profiles in the weld. While the 2D model is simplified and would benefit from further refinement and validation, preliminary results suggest that FEA could be used for residual stress measurement predictions. / by Dominic Solis. / S.B.

Thesis: S.M., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 2013. / Cataloged from PDF version of thesis. / Includes bibliographical references (pages 175-177). / The fluoride salt-cooled high-temperature reactor (FHR) combines high-temperature coated-particle fuel with a high-temperature salt coolant for a reactor with unique market and safety characteristics. This combination can eliminate large-scale radionuclide releases by avoiding major fuel failure during a catastrophic Beyond Design Basis Accident (BDBA). The high-temperature core contains liquid salt coolant surrounded by a liquid salt buffer; these salts limit core heatup while decay heat drops. The vessel insulation is designed to fail during a BDBA. The silo contains a frozen BDBA salt designed to melt and surround the reactor vessel during a major accident to accelerate heat transfer from the vessel. These features provide the required temperature gradient to drive decay heat from core to the vessel wall and to the environment below fuel failure temperatures. A 1047 MWth FHR was modeled using the STAR-CCM+ computational fluid dynamics package. Peak temperatures and heat transfer phenomena were calculated, focusing on feasibility of melting the BDBA salt that improves heat transfer from vessel to silo. A simplified wavelength-independent radiation model was examined to approximate the heat transfer capability with radiation heat transfer. The FHR BDBA system kept peak temperatures below the fuel failure point in all cases. Reducing the reactor vessel-silo gap size minimized the time to melt the BDBA salt. Radiation heat transfer is a dominant factor in the high-temperature accident sequence. It keeps peak fuel temperatures hundreds of degrees lower than with convection and conduction only; it makes higher core powers feasible. The FHR's atmospheric pressure design allows a thin reactor vessel, ensuring the high accident temperatures reach the vessel's outer surface, creating a large temperature difference from the vessel to the frozen salt. This greatly accelerates the heat transfer over current reactor designs with thick, relatively cool accident outer vessel temperatures. The frozen BDBA salt in the FHR places a limit on the upper temperature at the vessel outer boundary for significant time; it is a substantial heat sink for the accident duration. Finally, surrounding the FHR vessel, the convection of hot air, and circulating salt later in the accident, preferentially transports heat upward in the FHR; this provides a conduction path through the concrete silo to the atmosphere above the FHR. / by Matthew J. Minck. / S.M.

Thesis: Ph. D., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 2015. / This electronic version was submitted by the student author. The certified thesis is available in the Institute Archives and Special Collections. / "September 2015." Cataloged from student-submitted PDF version of thesis. / Includes bibliographical references (pages 273-289). / This work has focused on evaluating the thermo-mechanical performance of two innovative nuclear fuel designs: the ThO2-based resource-renewable boiling water reactor (RBWR) and silicon carbide (SiC) cladding in pressurized water reactors (PWRs). Since these designs employ non-traditional fuel rod materials and/or operating conditions, the fuel behavior and limits lie outside of current experience and must be appropriately assessed. The RedTail fuel performance code was developed to support early-stage innovative fuel designs using established models within a modular code structure. Additionally, the applied assumptions have been selected to allow for fast run times. This versatility and speed allow the RedTail fuel performance code to be applied towards broad scoping studies of new fuel rod concepts. With its use of (ThUPu)O2 and high cladding fluence, new physical models were applied to describe the ThO2-based RBWR fuel rods. The effects of lattice strain induced phonon scattering and semi-conductive electron transport on the thermal conductivity for this ternary fuel mixture were modeled. A new model has also been created to reconstruct the radial distribution of the fission rate in the fuel. Lastly, the effects of high fast neutron fluence on the accelerated oxidation and hydrogen pickup by Zircaloy-2 have been investigated and incorporated. In developing the ThO2-based RBWR concept, three designs have been proposed based on different goals and applied assumptions. With the exception of hydrogen uptake, none of the predicted parameters exceeded their limitations. However, the application of hydrogen-based cladding accident limits was found to be restricting. A sensitivity study revealed that in order to retain non-zero accident limits, an advanced cladding is required. Because of its lower corrosion rate and high strength in high temperature steam, composite ceramic SiC cladding has the potential to improve the response to severe accidents. Applicable SiC material properties have been investigated, and new swelling and thermal conductivity models have been applied to treat radiation damage effects. In addition, Weibull failure mechanics are applied to CVD SiC. The fuel performance of every SiC clad pin in a PWR core has been evaluated to find the core-wide leaker risk. To account for uncertainty in potential SiC cladding designs 24 separate cases were evaluated along with a comparison to ZIRLO®. The three-layer SiC architecture is found not to be a viable option due to large shutdown stresses. The two-layer design, however, shows much lower failure risk. This study indicates that significant opportunities exist to optimize core design to improve SiC cladding performance. / by Alexander James Mieloszyk. / Ph. D.

Thesis (Ph. D.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2006. / Includes bibliographical references (p. 93-104). / Quantum computation requires the ability to efficiently control quantum information in the presence of noise. In this thesis, NMR quantum information processors (QIPs) are used to study noise processes that compromise coherent control, to develop useful techniques for detecting noise, and to explore effective noise-protection schemes. A quantum simulation of the quantum sawtooth map in the perturbative parameter regime is used to study the effects of experimental noise on quantum localization, a highly sensitive quantum interference phenomenon that depends on the coherence of the localized state. Experimental data and numerical simulations show that the decoherent noise known to act on the system is relatively inconsequential in this implementation of the map, and that incoherent noise is the biggest challenge to implementing localization. While many incoherent processes appear decoherent, there are important differences. The distribution functions underlying incoherent processes are either static or slowly varying and so the errors introduced by these distributions are refocusable. The influence of incoherent noise is further explored in an experimentally implemented entangling operation, where incoherence can be difficult to separate from decoherence. By studying the fidelity decay under the cyclic entangling map, the effects of incoherence are easily distinguished from decoherence in experimental data. Decoherence free subspaces (DFSs) provide some of the most efficient schemes of avoiding decoherence from noise sources with underlying symmetries. / (cont.) To achieve an internal Hamiltonian structure that naturally fits a DFS encoding over a well-defined Hilbert space, we employ liquid crystal solvents to partially align a four-proton spin system, reintroducing the spin-spin dipolar couplings. In these experiments, enhanced coherent control is achieved by encoding logical qubits in a DFS. Robust control sequences enable high fidelity control in the DFS even when the system Hamiltonian is known with some uncertainty. / by Michael Kevin Henry, Jr. / Ph.D.

Thesis: S.M., 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 160-164). / Silicon carbide (SiC) has been proposed as an alternative to zirconium alloys used in current light water reactor (LWR) fuel cladding because it exhibits superior corrosion characteristics, high-temperature strength, and a 1000°C higher melting temperature, all of which are important during a loss of coolant accident (LOCA). To improve the performance of SiC cladding, a multilayered architecture consisting of layers of monolithic SiC (mSiC) and SiC/SiC ceramic matrix composite (CMC) has been proposed. In this work, the mechanical performance of both the tubing and the endplug joint of two-layer SiC cladding is investigated under conditions associated with the LOCA. Specifically, SiC cladding mechanical performance is investigated after exposure to 1,400°C steam and after quenching from 1,200°C into either 100°C or 90°C atmospheric-pressure water. The samples consist of two-layer SiC, with an inner SiC/SiC CMC layer and an outer monolith SiC layer. The relationship between mechanical performance and sample architecture is investigated through ceramography and internal void characterization. The two-layered SiC cladding design offered an as-received failure hoop stress of about 600 MPa, with little strength reduction due to thermal shock, and the tube failure hoop stress remained above 200 MPa after 48 hour high-temperature steam oxidation. The cladding showed pseudo-ductile behavior and failed in a non-frangible manner. The designs investigated for joint strength offered as-received burst strength above 30 MPa, although the impact of thermal shock and oxidation showed possible dependence on architecture. Overall, the cladding showed promising accident-tolerant performance. Because the implementation of SiC is complicated by the need for an open gap and low plenum pressure, thorium-based mixed oxides (MOX) are a promising fuel for SiC cladding because they have higher thermal conductivity and lower fission gas release (FGR). Previous efforts at MIT have modified the FRAPCON code to include thorium MOX fuel. In this work, the fission gas release and thermal conductivity models of FRAPCON-3.4-MIT are validated against published data. The results of this validation indicate a need to update the FGR model, which was accomplished in this work. / by Gregory Welch Daines / S.M.

Thesis (Ph. D.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2012. / Cataloged from PDF version of thesis. "February 2012." / Includes bibliographical references (p. 179-184). / This study develops a fundamentally new discrete generalized multigroup energy expansion theory for the linear Boltzmann transport equation. Discrete orthogonal polynomials are used, in conjunction with the traditional multigroup representation, to expand the energy dependence of the angular flux into a set of flux moments. The leading (zeroth) order equation is identical to a standard coarse group solution, while the higher order equations are decoupled from each other and only depend on the leading order solution due to the orthogonality property of the discrete Legendre polynomials selected. This decoupling leads to computational times comparable to the coarse group calculation but provides an accurate fine group energy spectrum. A source update process is also introduced which provides improvement of integral quantities such as eigenvalue and reaction rates over the coarse group solution. An online energy recondensation methodology is proposed to improve traditional multilevel approach in reactor core simulations. Since the discrete generalized multigroup (DGM) method produces an unfolded flux with a fine group structure, this flux can then be used to recondense the coarse group cross-sections using the obtained core level fine group DGM solution, which can be done iteratively. Computational tests on light water reactors and high temperature reactors are performed. Results indicate that flux can fully converge to the fine group solution with a computational time less than that of standard fine group calculation as long as a flat angular flux approximation is used spatially. The recondensation concept is extended to a nonlinear energy acceleration form. The method can effectively accelerate the fine group calculation by providing a more accurate initial guess provided from a few iterations of DGM recondensation calculation. Computational results show that the computational time and number of transport sweeps of the accelerated algorithm are much less than those of corresponding standard fine group calculations. / by Lei Zhu. / Ph.D.

Thesis (Ph. D.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2012. / Cataloged from PDF version of thesis. / Includes bibliographical references (p. 222-227). / Sodium Fast Reactors are one of the three candidates of GEN-IV fast reactors. Fast reactors play an important role in saving uranium resources and reducing nuclear wastes. Conventional fast reactors rely on transuranic fuels from reprocessing facilities, which are not available in the U.S. Thus, deployment of fast reactors requires decoupling from reprocessing facilities. This motivates the design and deployment of Uranium Startup sodium Fast Reactors (USFR) on a once-through fuel cycle in order to facilitate the transition to fast reactors by reducing their plant costs and increase capacity factor. Three different fuel types including uranium carbide (UC), metal (UZr) and uranium oxide (UO₂) are investigated and analyzed using the ERANOS code for potential use in USFR designs. A key enabling factor is use of high-albedo MgO or Zr reflectors in place of fertile blankets to reduce uranium enrichment and improve non-proliferation resistance. The different compositions in different fuel types result in different neutronic performance. The softer spectrum and lower allowable fuel volume fractions of oxide fuel have shorter fuel cycle length due to reactivity constraints, whereas fast neutron fluence plays an important role in determining the fuel cycle length in metal cores due to the harder spectrum. Moderators are deliberately added in the metal fuels to lower the fast neutron fluence. Carbide cores have a slightly harder neutron spectrum than oxide cores and a larger achievable fuel volume fraction. USFRs using all three fuel types (UC, U0 2 and UZr) have lower fuel cycle cost (6.27, 6.09 and 5.77mills/kWhe) and comparable uranium consumption (0.50, 0.55, and 0.53kgNatU/MWde) compared with typical LWRs (6.39nills/kWhe and 0.53kgNatU/MWde). All USFR designs have maximum neutron fluence below 5E23n/cm². All three USFR designs have pressure drop below 0.7MPa and maximum temperature below the limit for each fuel type. Both carbide and metal fuel have excellent passive safety performance. It is concluded that the USFR approach is a competitive way to accelerate fast reactor development. / by Tingzhou Fei. / Ph.D.