Thesis (Ph. D.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, February 2011. / "October 2010." Cataloged from PDF version of thesis. / Includes bibliographical references (p. 194-201). / An investigation into the properties and performance of a novel silicon carbide-based fuel rod cladding under PWR conditions was conducted. The novel design is a triplex, with the inner and outermost layers consisting of monolithic SiC, while the middle layer consists of a SiC fiberwound composite. The goal of this work was evaluation of the suitability of this design for use as a fuel rod cladding material in PWRs and the identification of the effects of design alternatives on the cladding performance. An in-core loop at the MITR-II was used to irradiate prototype triplex SiC cladding specimens under typical PWR temperature, pressure, and neutron flux conditions. The irradiation involved about 70 specimens, of monolithic as well as of triplex constitution, manufactured using several different processes to form the monolith, composite, and coating layers. Post-irradiation examination found some SiC specimens had acceptably low irradiation-enhanced corrosion rates and predictable swelling behavior. However, other specimens did not fare as well and showed excessive corrosion and cracking. Therefore, the performance of the SiC cladding will depend on appropriate selection of manufacturing techniques. Hoop strength testing found wide variations in tensile strength, but patterns or performance similar to the corrosion tests. The computer code FRAPCON, which is widely used for today's fuel assessment, modified properly to account for SiC properties, was applied to simulate effects of steady-state irradiation in an LWR core. The results demonstrated that utilizing SiC cladding in a 17x17 fuel assembly for existing PWRs may allow fuel to be run to somewhat higher burnup. However, due to lack of early gap closure by creep as well as the lower conductivity of the cladding, the fuel will experience higher temperatures than with zircaloy cladding. Several options were explored to reduce the fuel temperature, and it was concluded that annular fuel pellets were a solution with industrial experience that could improve the performance sufficiently to allow reaching 40% higher burnup. Management of the fuel-cladding gap was identified as essential for control of fuel temperature and PCMI. SiC cladding performance may be limited unless cladding/fuel conductivity or gap conductance is improved. / by David Michael Carpenter. / Ph.D.

Thesis (S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2007. / Includes bibliographical references (p. 45-46). / A time-dependent reliability evaluation of a two-loop passive Decay Heat Removal (DHR) system was performed as part of the iterative design process for a helium-cooled fast reactor. The system was modeled using RELAP5-3D. The uncertainties in input parameters were assessed and were propagated through the model using Latin Hypercube Sampling. An important finding was the discovery that the smaller pressure loss through the DHR heat exchanger than through the core would make the flow to bypass the core through one DHR loop, if two loops operated in parallel. This finding is a warning against modeling only one lumped DHR loop and assuming that n of them will remove n times the decay power. Sensitivity analyses revealed that there are values of some input parameters for which failures are very unlikely. The calculated conditional (i.e., given the LOCA) failure probability was deemed to be too high leading to the identification of several design changes to improve system reliability. This study is an example of the kinds of insights that can be obtained by including a reliability assessment in the design process. It is different from the usual use of PSA in design, which compares different system configurations, because it focuses on the thermal-hydraulic performance of a safety function. / by Francisco J. Mackay. / 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. 69-70). / PURPOSE: Prompt gamma rays emitted from proton-nucleus interactions in tissue present a promising non-invasive, in situ means of monitoring proton beam based radiotherapy. This study investigates the fluence and energy distribution of prompt gamma rays emitted during proton irradiation of phantoms. This information was used to develop a correlation between the measured and calculated gamma emission and the proton beam range, which would allow treatments to more effectively exploit the sharp distal falloff in the dose distributions of protons. METHOD & MATERIALS: A model of a cylindrical Lucite phantom with a monoenergetic proton beam and an annular array of ideal photon tallies arranged orthogonal to the beam was developed using the Monte Carlo code MCNPX 2.6.0. Heterogeneous geometries were studied by inserting metal implants into the Lucite phantom, and simulating a phantom composed of bone and lung equivalent materials and polymethyl methacrylate. RESULTS: Experimental and computational results indicated a correlation between gamma emission and the proton depth-dose profile. Several peaks were evident in the calculated energy spectrum and the 4.44 MeV emission from 12C was the most intense line having any apparent correlation with the depth dose profile. Arbitrary energy binning of 4-5 MeV and 4-8 MeV was performed on the Monte Carlo data; this binned data yielded a distinct emission peak 1cm proximal to the Bragg peak. In all cases in the Lucite phantom the position of the Bragg peak's 80% distal falloff corresponded with the position of the 4-8MeV binned 50% distal falloff. The 4-5MeV binning strategy was successful with the heterogeneous phantom in which the proton beam entered lung and stopped in bone. However, the density disparity between the bone and lung equivalent materials rendered this technique unsuccessful for the heterogeneous phantom in which the beam entered bone and stopped in lung. For this 1.4MeV binning was conducted, assessing the 1.37 MeV characteristic gamma peak of 24Mg, which was only present in the lung slab. CONCLUSIONS: The results are promising and indicate the feasibility of prompt gamma emission detection as a means of characterizing the proton beam range in situ. This study has established the measurement and omputational tools necessary to pursue the development of this technique. / by John R. Styczynski. / S.M.and S.B.

Thesis (Ph. D.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2010. / Cataloged from PDF version of thesis. / Includes bibliographical references (p. 211-216). / The three-dimensional interaction of a magnetized, collisionless flowing plasma with a non-emitting conducting sphere is solved in the entire range of physically allowed parameters, in the ion-collecting regime. This can be considered as the "spherical Mach probe" problem, establishing how the ion flux to the surface varies with orientation and external velocity; the study is however of broader interest, as the sphere can also be seen as a dust particle or any ionospheric body. The core tool developed for this study is the fully parallelized (particle + field solver) Particle-In-Cell code SCEPTIC3D, three-dimensional evolution of SCEPTIC, accounting for the full ion distribution function and Boltzmann electrons. Investigations are first carried out in the quasineutral limit. Results include a report of ion current dependence on the external plasma parameters, as well as a theoretical calibration for transverse Mach probes with four electrodes oriented at 450 to the magnetic field in a plane of flow and magnetic field, valid for arbitrary temperature and ion magnetization. The analysis is preceded by an independent semi-analytic treatment of strongly magnetized ion collection by oblique surfaces, successfully validating SCEPTIC3D's behaviour. / (cont.) The finite shielding length regime is more complex, and an important transition in plasma structure occurs when the Debye length goes over the average ion Larmor radius. Studies of ion collection show that the ion current can exceed the (unmagnetized) OML limit at weak magnetization, and the Mach probe calibration method proposed in the context of quasineutral plasmas holds up to Debye lengths equal to about 10% of the probe radius. A further analysis consists in calculating the force exerted by the flow on spherical dust. In short Debye length plasmas a strong drag component antiparallel to the convective electric field forms, causing the dust to spin faster than what predicted by its Larmor frequency. At intermediate and large Debye length the ion-drag in the direction of transverse flow is found to reverse in subsonic conditions, but the internal Laplace force appears to be positive, and larger in magnitude than the negative ion-drag. / by Leonardo Patacchini. / Ph.D.

Thesis: S.M., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 2016. / Cataloged from student-submitted PDF version of thesis. / Includes bibliographical references (pages 80-82). / Zirconium-based alloys, which have been employed as cladding materials in nuclear industry, pick up hydrogen during service because of corrosion in water. This is one of the degradation processes that challenge the safe operation of nuclear reactors. Composition of commercial zirconium-based alloys has been developed experimentally, aiming to improve corrosion and hydrogen resistance. The empirical testing process is costly, time-consuming and can provide little knowledge on the underlying mechanisms. In order to understand the critical step of hydrogen entering zirconium-based alloys, we study hydrogen migration in monoclinic zirconium dioxide (m-ZrO2 ), the oxide phase, which is present in the outer protective oxide layer contacting the cooling water. First principles calculations are casted to study hydrogen migration at atom-istic scale. All possible stable sites for hydrogen to take in the structure are determined based on energy minimization. Barriers and saddle point configurations of elemental diffusion paths are obtained using the nudged elastic band method. Each extended diffusion path can be constructed from these elemental migration steps ... / by Ming Yang. / S.M.

Thesis: S.M., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 2014. / Cataloged from PDF version of thesis. / Includes bibliographical references (pages 147-149). / In this thesis, we investigated the Adiabatic, Theta, and Multigrid Amplitude Function (MAF) methods for solving the Method of Characteristics (MOC) formulation of the time-dependent neutron transport equation. The transient transport versions of the 2D LRA and C5G7 benchmarks were used to assess the performance and accuracy of these methods. We began by deriving the CMFD-accelerated MOC algorithm in 2D steady state form and examining the effects of various MOC and CMFD parameters on eigenvalue convergence. The C5G7 problem showed similar acceleration performance for 2, 4, and 7 CMFD energy group structures. CMFD meshes at or near the pin-cell level resulted in the greatest speedups of 15-45x in run time and 30-240 x in number of MOC iterations for both problems. A relaxation factor on the nonlinear diffusion coefficient was required to maintain stability for both problems with optimum values between 0.4-0.7. A sensitivity study was conducted on the C5G7 and LRA transient problems to understand the effects of time step and spatial mesh sizes on the solution accuracy and run time performance. The shape function time step size had a large effect on the solution accuracy for the MAF and Theta methods in solving the LRA problem. All methods showed moderate sensitivity to the amplitude function step size, where increasing step size shifted the peak power to earlier times. The coarse mesh size did not have a significant effect on solution accuracy, but meshes on the pin-cell level were clearly preferred to reduce run time. The overall run time performance between the three methods was mixed. The MAF and Theta methods displayed ~84% speedup over the Adiabatic method for the LRA problem, but all methods had similar run times for the C5G7 problem. This inconsistency is likely due to the more drastic flux shape change during the LRA transient and the ability of the MAF and Theta methods to more accurately treat the flux shape temporal derivative. These results demonstrate that the Adiabatic, Theta, and MAF methods are computationally efficient methods for solving the time-dependent neutron transport equation and warrant further investigation. There are clear advantages to each method and the optimal method will likely depend on the transient characteristics of the problem being studied. / by Samuel Christopher Shaner. / S.M.

Thesis: Ph. D., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 2017. / 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 295-307). / An important Fluoride-salt-cooled High-temperature Reactor (FHR) development step is to design, build, and operate a test reactor. Through a literature review, liquid-salt coolant thermophysical properties have been recommended along with their uncertainties of 2-20%. This study tackles determining the effects of these high uncertainties by proposing a newly developed methodology to incorporate uncertainty propagation in a thermal-hydraulic safety analysis for test reactor licensing. A hot channel model, Monte Carlo statistical sampling uncertainty propagation, and limiting safety systems settings (LSSS) approach are uniquely combined to ensure sufficient margin to fuel and material thermal limits during steady-state operation and to incorporate margin for high uncertainty inputs. The method calculates LSSS parameters to define safe operation. The methodology has been applied to two test reactors currently considered, the Chinese TMSR-SF1 pebble bed design and MIT's Transportable FHR prismatic core design; two candidate coolants, flibe (LiF-BeF2) and nafzirf (NaF-ZrF4); and forced flow and natural circulation conditions to compare operating regions and LSSS power (maximum power not exceeding any thermal limits). The calculated operating region accounts for uncertainty (2 [sigma]) with LSSS power (MW) for forced flow of 25.37±0.72, 22.56±1.15, 21.28±1.48, and 11.32±1.35 for pebble flibe, pebble nafzirf, prismatic flibe, and prismatic nafzirf, respectively. The pebble bed has superior heat transfer with an operating region reduced ~10% less when switching coolants and ~50% smaller uncertainty than the prismatic. The maximum fuel temperature constrains the pebble bed while the maximum coolant temperature constrains the prismatic due to different dominant heat transfer modes. Sensitivity analysis revealed 1) thermal conductivity and thus conductive heat transfer dominates in the prismatic design while convection is superior in the pebble bed, and 2) the impact of thermophysical property uncertainties are ranked in the following order: thermal conductivity, heat capacity, density, and lastly viscosity. Broadly, the methodology developed incorporates uncertainty propagation that can be used to evaluate parametric uncertainties to satisfy guidelines for non-power reactor licensing applications, and method application shows the pebble bed is more attractive for thermal-hydraulic safety. Although the method was developed and evaluated for coolant property uncertainties for FHR, it is readily applicable for any parameters of interest. / by Rebecca Rose Romatoski. / Ph. D.

Thesis (Ph. D.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2007. / "June 2007." In title on t.p. "[tau]" appears as the lower-case Greek letter; "e" appears subscript; "[beta]" appears as the lower case Greek letter and the superscript minus symbol appears over the Greek letter beta. / Includes bibliographical references (leaves 135-140). / The Levitated Dipole Experiment (LDX) is a novel experiment to study the confinement of a high-temperature plasma in the magnetic field of a superconducting ring of wire. The levitated magnet produces a poloidal closed-line magnetic field characteristic of an ideal point dipole or a hard core Z-pinch magnetic configuration. The point dipole and hard core Z-pinch configurations share similar physics and may be respectively considered to be the zero and large aspect ratio approximations to LDX. The present work focuses on a hard-core Z-pinch magnetic configuration. An analysis is presented that theoretically predicts (1) the maximum pressure p., (2) the energy confinement time TE and (3) the average beta / by solving a proposed self-consistent model of plasma. The model makes the optimistic assumption that transport is purely classical in the region of the profile that is magnetohydrodynamically (MHD) stable against interchange modes. For the interchange unstable region, a quasilinear MHD transport model is developed. The analysis of MHD quasilinear transport starts with an assessment of stability corrections due to axial flows. The axial flows are taken as an approximation to the LDX toroidal flows, expected to appear due to non-ambipolar transport. It is shown that the subsonic axial flows create only negligible correction to the plasma stability and the MHD transport analysis is performed for a static plasma. The evolution of the particle density, energy and magnetic field in the MHD unstable region is investigated using the quasilinear approximation. The exact transport equations are derived for a static plasma in the hard core Z-pinch magnetic configuration. The equations are generalized to an arbitrary axisymmetric closed-filed line magnetic configuration. / (cont.) It is shown that violation of the marginal stability criterion leads to a rapid time-scale transport (i.e. much faster than classical transport), which brings the pressure profile back to marginal stability and forces particle density to be inversely proportional to V = d / B. The applicability of the quasilinear approximation is numerically tested in a hard core Z-pinch magnetic configuration using a non-linear numerical code. The numerical results confirm the theoretical conclusions that the plasma maintains its marginally stable pressure profile through anomalous transport. The requirement of the marginally stable pressure profile plus p - V-'density profile completes the model and provides sufficient information to calculate TE and /3 in the hard core Z-pinch magnetic configuration. The predictions show that the performance of a large-aspect ratio LDX is strongly coupled to the maximum achievable edge temperature with relatively good performance achieved when T, > 10 eV. Performance should be further improved by the finite aspect ratio in the real experiment. Analytic and numerical calculations lead to explicit scaling relations for TE and , that can be tested in future LDX experiments. / by Alexei Kouznetsov. / Ph.D.

Thesis (S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2006. / Includes bibliographical references (p. 589). / Stress Corrosion Cracking (SCC) in and around primary loop piping welds in Boiling Water Reactors has been observed worldwide as plants continue to operate at temperatures and pressures near 2880C (5500F) and 6.9 MPa (1000 PSI). An experimental program was designed to explore the effect of thermal aging on the SCC crack growth rate in weld materials for type 316 and 304 stainless steels. An autoclave facility was designed and constructed for the measurement of SCC crack growth rates under BWR conditions and testing was underway at the time of this writing. The effects of composition and thermal aging on mechanical properties (i.e. tensile, micro-hardness, nano-hardness, Jic, and Charpy-impact toughness) was in process and initial results show an increase in yield strength and a decrease in fracture toughness after aging for 1000 hours at 430 and 400 *C. Thermal aging results in no discernable changes to the 6-ferrite morphology when viewed optically at 500 X agnifications and in the scanning electron microscope. / by Jeff Hixon. / S.M.

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. 445-466). / A high-speed video and infrared thermography based technique has been used to obtain detailed and fundamental time- and space-resolved information on pool boiling heat transfer. The work is enabled by recent advances in heat transfer diagnostics and motivated by increased interest in the enhancement of heat transfer for applications such as micro-electronics, space heat-transfer systems, and advanced nuclear reactors. This study critically examined the fundamental processes occurring during nucleate boiling, critical heat flux, and rewetting on thin-film heating elements. A significant focus of the work was to investigate and explain the modification of these heat transfer phenomena through the addition of silica and diamond nanoparticles to the working fluid. Bubble departure diameter and frequency, growth and wait times, and nucleation site density were measured for every nucleation site during nucleate pool boiling at multiple superheats. The data were compared with decades-old and poorly-validated models and correlations, and were used to evaluate the relative contribution of the superheated liquid layer and microlayer evaporation to bubble growth. Deterioration in nucleate boiling heat transfer of water-based nanofluids was observed. It was determined that a reduction in the static contact angle, caused by nanoparticle deposition on the surface during boiling, created a larger energy barrier for nucleation, which in turn reduced bubble departure frequency and nucleation density, thus resulting in a reduced heat transfer coefficient. / (cont.) Critical heat flux enhancement in nanofluids of up to 100% was experimentally observed. The cause of this enhancement was determined to be the decreased static contact angle of nanofluid boiled surfaces. The increased wettability modified the growth of bubbles prior to CHF and promoted rewetting of hotspots at CHF. In parallel quenching tests, rewetting temperatures and velocities were simultaneously measured for the first time. Surfaces that had been pre-boiled in nanofluids were found to have significantly higher rewetting temperatures and velocities than clean surfaces. Interpretation of the experimental data was conducted with consideration of the governing surface parameters and existing models. It was found that there is significant room for improvement of most pool boiling models, especially with regard to surface effects. The research performed in this thesis help demonstrate the power of the infrared thermography technique and its potential for future improvement of boiling models. / by Craig Douglas Gerardi. / Ph.D.