Thesis: Ph. D., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 2014. / Cataloged from PDF version of thesis. / Includes bibliographical references (pages 409-425). / Heat transfer in high aspect ratio mini-channels has important applications for materials test reactors using plate-type fuel. These fuel plates typically possess coolant channels with hydraulic diameters on the order of 4 mm or less. The single and two-phase heat transfer in such channels has not been well-characterized, especially in regard to the onset of nucleate boiling. While surface effects are known to dramatically influence the incipience of boiling, they have not been widely considered under forced convection. Since the limiting safety system setting for the MITR is the onset of nucleate boiling, there is considerable interest in better characterizing the phenomenon in such channels. This study presents a first-of-a-kind, two-phase flow facility designed to measure the single-phase heat transfer coefficient and onset of nucleate boiling in a high aspect ratio mini-channel over a wide range of flow conditions while also permitting high speed visualization of the entire surface. The single-phase heat transfer coefficient is measured for mass fluxes ranging from 750 kg/m2-sec up to 6000 kg/m2 -sec and for sub-coolings ranging from 20 °C to 70 °C. The onset of nucleate boiling superheat and heat flux are measured for mass fluxes ranging from 750 kg/m2- sec to 3000 kg/m2-sec and for sub-coolings ranging from 10 °C to 45 °C. Measurements are supported with high speed videography to visualize bubble incipience when conditions permit. The influence of surface wettability on the incipience point is also investigated by performing tests on a surface oxidized at high temperature in air. Using a boundary layer analysis along with experimental data obtained in the study, a semi-analytical correlation is developed to predict the single-phase heat transfer coefficient in high aspect ratio rectangular channels. The correlation accounts for effects from secondary flows and heating asymmetry, and is suitable for both the transition and fully turbulent flow regimes. The new correlation predicts the Nusselt number with a mean absolute error of 4.9% in the range of 2.2<Pr<5.5 and 4000<Re<70,000. The onset of nucleate boiling heat flux on the reference surface is adequately predicted with the correlation of Bergles and Rohsenow, as long as the appropriate single-phase heat transfer prediction is used. However, the oxidized surface displays a modest increase in the incipient heat flux due to the improved wettability. This effect is not captured in the correlation of Bergles and Rohsenow, but is accounted for in other correlations such as that of Davis and Anderson. Surface science measurements are presented for prototypical materials test reactor fuel surfaces to quantify effects of roughness, oxidation, and surface contamination on wettability. Overall, surface cleanliness is found to have a profound effect on wettability, and in turn, is expected to influence boiling incipience. / by Eric Christopher Forrest. / Ph. D.

Thesis (Ph. D.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2007. / Includes bibliographical references (p. 283-289). / Passive cooling via natural circulation of gas after a loss of coolant (LOCA) accident is one of the major goals of the Gas-cooled Fast Reactor (GFR). Due to its high surface heat flux and low coolant velocities under natural circulation in post-LOCA scenarios, the capability of turbulent gas flow to remove heat from the GFR core can be impaired by either a buoyancy effect or an acceleration effect. These phenomena lead to a Deteriorated Turbulent Heat Transfer (DTHT) regime. To predict accurately the cladding temperature at the hot spot, reliable heat transfer correlations that account correctly for these effects are needed. This work addresses this need by experimentally obtaining heat transfer data and developing new heat transfer correlations that can be used in system analysis codes, such as RELAP5-3D, to reduce uncertainties of predictions in these DTHT regimes. An experimental facility was designed and built using similitude analysis to match key experimental loop parameters to the GFRs' Decay Heat Removal (DHR) system operating conditions to the largest extent possible. Through a thorough literature survey two non-dimensional numbers namely (1) the buoyancy parameter (Bo*) and (2) the acceleration parameter (K,) were identified as important indicators of the DTHT regime. The experimental data was collected for a range of (1) inlet Reynolds number from 1800 to 42,700, (2) inlet Bo* up to 1X10-5 (3) and inlet Kv up to 5x10-6. The data showed significantly higher reduction of the Nusselt number (up to by 70%) than previously reported (up to 50%). Also, the threshold at which DTHT regime occurs was found to be at smaller non-dimensional numbers than previously reported. A new phenomenon "re-turbulization", where the laminarized heat transfer recovers back to turbulent flow along the channel, was observed in the experiment. / (cont.) A new single phase gas flow heat transfer map is proposed based on the non-dimensional heat flux and the Reynolds number in our data, and is shown to compare well with data in the literature. Three sets of new correlations were developed, which reflect both the buoyancy and acceleration effects and have better accuracy as well as ease of numerical implementation than the existing correlations. The correlations are based on the Gnielinski correlation and replace the Reynolds number subtracting constant by a functional form that accounts for the buoyancy and acceleration effects separately, or in the combined form through a newly introduced non-dimensional "DTHT" number. The three correlation types have different complexity level, with the first being the most complex and the third being the most simple and easy to apply without any need for iterations. Additional runs with natural circulation showed that the friction factor in the DTHT regime could be significantly higher than predicted by conventional friction factor correlations, although more experiments will be needed to develop reliable correlations for pressure drop in these regimes. Overall, it is concluded that due to the low heat transfer coefficient and increased friction factor in the DTHT regime, the GFR DHR system should be ideally designed to operate outside the DTHT regime to (1) avoid reduction of heat transfer capability, (2) avoid increase of pressure drop, and (3) reduce uncertainties in predictions of the cladding temperature. / by Jeongik Lee. / Ph.D.

Thesis: S.M., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 2014. / Cataloged from PDF version of thesis. / Includes bibliographical references (pages 197-203). / Over the past 20 years, parallel computing has enabled computers to grow ever larger and more powerful while scientific applications have advanced in sophistication and resolution. This trend is being challenged, however, as the power consumption for conventional parallel computing architectures has risen to unsustainable levels and memory limitations have come to dominate compute performance. Multi-core processors and heterogeneous computing platforms, such as Graphics Processing Units (GPUs), are an increasingly popular paradigm for resolving these issues. This thesis explores the applicability of shared memory parallel platforms for solving deterministic neutron transport problems. A 2D method of characteristics code - OpenMOC - has been developed with solvers for shared memory multi-core platforms as well as GPUs. The multi-threading and memory locality methodologies for the multi-core CPU and GPU solvers are presented. Parallel scaling results using OpenMP demonstrate better than ideal weak scaling and nearly perfect strong scaling on both Intel Xeon and IBM Blue Gene/Q architectures. Performance results for the 2D C5G7 benchmark demonstrate up to 50x speedup for MOC on a GPU. The lessons learned from this thesis will provide the basis for further exploration of MOC on many-core platforms and GPUs as well as design decisions for hardware vendors exploring technologies for the next generation of machines for scientific computing. / by William Robert Dawson Boyd III. / S.M.

Thesis (S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2006. / Includes bibliographical references (leaves 113-115). / The ACR-700 is an advanced pressure-tube (PT) reactor being developed by Atomic Energy of Canada Limited (AECL). As in conventional CANDU reactors, the PTs are horizontal. Each PT is surrounded by a calandria tube (CT), and the gap in between is filled with carbon dioxide gas. The space between the CTs is filled with the heavy-water moderator. One postulated accident scenario for ACR-700 is the complete coolant flow blockage of a single PT. The flow is not monitored within each individual PT, thus during the early stages of this accident the reactor remains at full power and full pressure, resulting in rapid coolant boil-off and fuel overheating. Melting of the Zircaloy (Zry) components of the fuel bundle (cladding, end plates and end caps) can occur, with relocation of some molten material to the bottom of the PT. The hot spot caused by the molten Zry/PT interaction may cause PT/CT failure due to localized plastic strains. Failure of the PT/CT results in depressurization of the primary system, which triggers a reactor scram, after which the decay heat is removed via reflooding, thus PT/CT rupture effectively terminates the accident. Clearly, prediction of the time scale and conditions under which PT/CT failure occurs is of great importance for this accident. We analyzed the following key phenomena occurring after the blockage: (a) Coolant boil-off (b) Cladding heat-up and melting (c) Dripping of molten Zircaloy (Zry) from the fuel pin (d) Thermal interaction between the molten Zry and the PT (e) Localized bulging of the PT (f) Interaction of the bulged PT with the CT Simple one-dimensional models were adequate to describe (a), (b) and (c), while the three-dimensional nature of (d), (e) and (f) required the use of more sophisticated models including a finite-element description of the thermal transients within the PT and the CT, implemented with the code COSMOSM. / (cont.) The main findings of the study are as follows: (1) Most coolant boils off within 3 s of accident initiation. (2) Depending on the magnitude of radiation heat transfer between adjacent fuel pins, the cladding of the hot fuel pin in the blocked PT reaches the melting point of Zry in 7 to 10 s after accident initiation. (3) Inception of melting of the UO2 fuel pellets is not expected for at least another 7 s after 2Zry melting. (4) Several effects could theoretically prevent molten Zry dripping from the fuel pins, including Zry/UO2 interaction and Zry oxidation. However, it was concluded that because of the very high heat-up rate typical of the flow blockage accident sequence, holdup of molten Zry would not occur. Experimental verification of this conclusion is recommended. (5) Once the molten Zry relocates to the bottom of the PT, a hot spot is created that causes the PT to bulge out radially under the effect of the reactor pressure. The PT may come in contact with the CT, which heats up, bulges and eventually fails. / (cont.) The inception and speed of the PT/CT bulging and ultimately the likelihood of failure depend strongly on the postulated mass of molten Zry in contact with the PT, and on the value of the thermal resistance at the Zry/PT interface. It was found that a Zry mass =/< 10 g will not cause PT/CT failure regardless of the contact resistance effect. On the other hand, a mass of 100 g would be sufficient to cause PT/CT failure even in the presence of a thick 0.2 mm oxide layer at the interface. The characteristic time scales for this 100-g case are as follows: PT bulging starts within 3 s of Zry/PT contact - PT makes contact with the CT in another 2 s - CT bulging starts in less than 1 s - CT failure occurs within another 5 s. Thus, the duration of the PT/CT deformation transient is 11 s, which gives a total duration of the accident (from PT blockage to PT/CT failure) of 18 to 21 s. / by Craig Douglas Gerardi. / 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. 67). / A simple model for simulating deuterium tritium burn in inertial confinement fusion capsules is developed. The model, called the Isothermal Rarefaction Model, is zero dimensional (represented as ordinary differential equations) and treats disassembly in the isothermal limit. Two substantive theoretical developments are contained in this model; one is an improved treatment of fast alpha slowing down, and the other is a calculation of the fusion product source distributions and their energy moment. The fast alpha stopping treatment contains a derivation of the Fraley fractional energy splitting functional form, fe = 1/(1 + xTe), resulting in an expression for the numerical factor x which will be defined as the Fraley parameter. The average thermal energy which is lost from the thermal ion distribution when two particles fuse is found from the energy moment of the fusion product source distribution. This energy contributes to the energy of the fusion products. A third theoretical development that is discussed for completeness and future use, but not yet incorporated in the Isothermal Rarefaction Model, is the 4T theory of matter-radiation energy exchange in homogenous optically thick media. The isothermal rarefaction model assumes an optically thin to marginally thick plasma, and only Bremsstrahlung emission and absorption are treated in this thesis. The 4T theory for optically thick media has been published. A sampling of results using the Isothermal Rarefaction Model is presented. / by Conner Daniel Galloway. / S.M.and S.B.

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 [183]-190). / Multiphase computational fluid dynamics (M-CFD) modeling approaches allow for the prediction of critical three-dimensional thermal-hydraulics phenomena in nuclear reactor applications. The advancement and consistent adoption of such tools could transform the industry's approach to the design of reliable systems, and the efficient operation of systems existing, which in the past have been dependent upon correlation-based sub-channel analysis codes. The success of these M-CFD methods in simulating two-phase flow and boiling heat transfer depends on their demonstrated accuracy and robustness, which signals a dual need for the comprehensive analysis of existing data and a reevaluation of the underlying physics. By virtue of the Eulerian-Eulerian two-fluid approach, additional terms manifest in the M-CFD phase momentum equations, which represent information that has been lost, and require closure through prescription of interfacial force models. These momentum "closures" are vital to M-CFD prediction of mean flow profiles, including void fraction and phase velocity distributions, and require high-resolution, robust models to perform accurately throughout a diverse array of flow conditions. While an overwhelming number of models have been developed with a wide range of varying performance, no consensus exists about how to assemble these models successfully in a CFD framework, suggesting that their predictive power is still limited. The lift force, responsible for lateral void fraction redistribution, is particularly refractory to the development of a consistent modeling strategy for these closures. Historically, CFD approaches have been forced to inconsistently leverage existing models derived for single bubbles in laminar flow, which disregard the complex dynamics and interactions of bubbles with turbulence and bubble wakes. Current understanding of the lift force in turbulent flow has been limited to qualitative evidences that small spherical bubbles experience a positive lift, resulting in a wall-peaked void fraction distribution, while larger deformed bubbles experience a negative lift and corresponding center-peaked profile. The present work brings forward a new physical interpretation of the lift force in turbulent bubbly flow through a synthesis of information from DNS studies, fine and coarse scale experiments, and analytical investigations. To overcome the limitations of previous models, a simple dimensionless quantity, the Wobble number, a number which systematically describes the unsteady behavior of bubbles in turbulent flow conditions, is proposed. Introducing this dependency into the lift formulation allows for precise identification of lift inversion, which alone exceeds capabilities of existing models. Additionally, the model is extended to account for group behavior with the introduction of a swarm-like function based on void fraction. Its formulation is built on the conceptual understandings of drift phenomena, bubble interaction probability, and the maximum packing factor for dispersed bubbly flow. These two key mechanisms are assembled into a lift model for turbulent bubbly flow, which is implemented in CFD and validated on several experimental databases spanning an extensive range of flow conditions. Error analyses demonstrate the new formulation's robustness and predictive abilities, allowing for a more comprehensive representation of the two-phase phenomena particularly significant in nuclear reactor applications; moreover, it avoids the introduction of case-specific adjustments to unphysical coefficients and tunable parameters which are characteristic and typical limitations of previous models, indicating another valuable improvement. Finally, the new model's performance in a prototypical rod bundle is evaluated and a qualitative assessment of its applicability in a nuclear reactor geometry context is demonstrated. / by Rosemary M. Sugrue. / Ph. D.

Thesis (S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2008. / "May 2008." / Includes bibliographical references (p. 60-62). / As part of a multi-scale approach for modeling hydrogen embrittlement in hardened steels we have investigated, employing density functional theory methods, the stability and concentrations of the point defect clusters present in metastable Fe-C-H alloys with vacancies. The defect spectrum is found to be strongly dependent on the local vacancy concentration, and for low hydrogen levels sharp highly non-linear changes in the defect cluster population are observed at critical vacancy concentrations. This critical-like behavior suggests an energy activation mechanism which can be characterized by an effective defect-cluster formation energy barrier. By analogy with similar activated processes such as the liquid-to-glass transition in super-cooled liquids, we postulate that this criticality is associated with the presence of deep wells in the energy landscape where chemical composition plays the role of generalized coordinate. Increases in the hydrogen content have the qualitative effect of reducing the slopes in the defect concentrations. The drastic sensitivity of the defect cluster spectrum to local changes in vacancy and impurity concentrations implies that in proximity of surfaces and extended defects multiple defect clusters become statistically significant and migration dependent phenomena, such as creep-relevant to hydrogen embrittlement-and super-diffusion, should be controlled by multiple activation barriers. / by Paul Rene Monasterio Velásquez. / S.M.

Thesis: S.M., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 2018. / 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 41-42). / Reactivity initiated accidents (RIAs) are a potential concern for nuclear reactor safety. In RIA scenarios, following the insertion of positive reactivity, e.g. by an unanticipated extraction of the control rods, the reactor power may increase exponentially. The period of the exponential rise, τ, depends on the amount of positive reactivity inserted, as well as the fuel composition. During such an event, boiling of the water coolant can provide not only an effective way of heat removal, but also a stabilizing, negative reactivity feedback. However, the reactor power could reach extremely high levels and lead to a boiling crisis, e.g. by departure from nucleate boiling (DNB), in turn leading to fuel damage. The aim of the current work is to improve the understanding of transient DNB phenomena. This goal was achieved by running experiments on a specially designed flow boiling platform, which includes high speed video (HSV) and high speed infrared (HSIR) diagnostics. Specifically, the IR radiation recorded by the HSIR camera was analyzed with dedicated post processing algorithms that enable measurements of the time-dependent temperature and heat flux distributions on the boiling surface. Experiments were performed on a flat heater in upward flow conditions at atmospheric pressure. This work explores the effects of flow velocity, liquid subcooling and exponential power escalation period on critical heat flux (CHF). The results show that, for our flow conditions, the CHF value does not depend on the escalation period for periods longer than 100 ms, and is essentially the same as in steady-state boiling. For shorter periods, CHF increases as the escalation period decreases, and the effect of flow velocity becomes less important at short periods. Larger subcooling was shown to increase the CHF at all conditions. For extreme cases of 50 K and 75 K of subcooling the entire heating surface was covered by tiny bubbles. Those bubbles had a very short (less than 50 [mu]s) lifetime and were quenched right after the nucleation. Such behavior prevented bubbles from coalescing and resulted in a very efficient heat transfer mechanism. CHF was observed at much higher values compared to steady boiling conditions, when nucleation site density and bubble size were large enough for the bubble to start coalescing. An interesting effect was observed for very short periods (5, 10 and 20 ms) and low subcoolings (10 K). At those conditions the boiling surface experiences CHF during the growth of the first generation of bubbles. Therefore, the points of ONB and CHF are almost coincident, with CHF delayed only by the time required for adjacent bubbles to coalesce and the microlayer underneath them to evaporate. / by Artyom Kossolapov. / S.M.

Thesis: S.B., 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 73-76). / Advanced modelIng and sImulatIon (M&S) of nuclear systems could offer a key contrIbutIon In enhancIng the competItIveness and safety performance of nuclear power plants. Large multI-organIzatIonal InItIatIves such as the ConsortIum for Advanced SImulatIon of LIght Water Reactors (CASL) and the Nuclear Energy Advanced ModelIng and SImulatIon (NEAMS) emphasIzes the Importance of M&S research to the U.S. nuclear Industry. UncertaInty QuantIfIcatIon (UQ) represents a fundamental area of research necessary to expand the applIcatIon of M&S Into nuclear Industry, but the fIeld Is stIll not mature, and no general consensus exIsts on current UQ methods. In thIs study, the ASME V&V20 I a proposed methodology for UQ of CFD -- Is applIed to a benchmark nuclear system turbulent mIxIng case In an effort to assess the applIcabIlIty and lImItatIons of the standard. / by Andres Felipe Alvarez. / S.B.

Thesis: Ph. D., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 2015. / Cataloged from PDF version of thesis. / Includes bibliographical references (pages 463-478). / The comprehensive analysis was performed to study turbulent transport in Alcator CMod plasmas in this thesis. A new Correlation Electron Cyclotron Emission (CECE) diagnostic was designed and installed as a part of this thesis work. Using this diagnostic, we measured local T fluctuations in r/a >/~ 0.75 in C-Mod for the first time. This thesis work provided new information about the Ohmic confinement transition, from the linear to the saturated confinement regime with the increase in average density. It was found that Te fluctuations near the edge (r/a0~.85) tend to decrease across the Ohmic confinement transition. Although the Ohmic confinement transition has been considered predominantly as a result of the linear turbulence mode transition, we found no changes in the dominant turbulence mode across this transition via gyrokinetic analysis using the code, GYRO. The GYRO simulations performed near the edge reproduce experimental ion heat flux and Te fluctuations, but electron heat flux was under-predicted. Considering that both ion heat flux and the T fluctuations mainly come from ion scale turbulence, the under-prediction of electron heat flux suggests the importance of electron scale turbulence. Intrinsic rotation reversals in C-Mod plasmas were studied in this thesis. Similar changes in electron temperature fluctuations, the reduction of Te fluctuations near the edge, were observed across RF rotation reversals and Ohmic rotation reversals. The gyrokinetic and self-similarity analyses also showed similarities between rotation reversals in Ohmic and RF heated discharges. These observations suggest that the physics of Ohmic confinement transition and the rotation reversal can be applied to the physics of rotation reversal in RF heated discharges. This thesis also found the reduction of Te fluctuations inside pedestal region with the transition from low to high energy confinement regime, which indicates the changes in core turbulence are correlated with the global energy confinement. / by Choongki Sung. / Ph. D.