Thesis (Ph.D.)--Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 2000. / Includes bibliographical references (p. 187-191). / In this thesis a study of molybdenum sources and screening in Alcator C-Mod is presented. This work contributes in characterizing the performance of molybdenum, a high Z material, as a tokamak first wall surface. Understanding the molybdenum production mechanism is crucial if one wants to minimize it. A simple physical sputtering model has been developed that calculates the molybdenum source at the divertor, providing satisfactory agreement with the spectroscopic results. The effect of deuterons, boron ions, and redeposited molybdenum incident on the target is included in the calculation which shows that the boron ions dominate the sputtering. It is also found that the probability of molybdenum being "promptly" redeposited (within a gyration after having been sputtered) can be as high as 80%. High probability of redeposition is favorable because it reduces the net erosion. Specifically, it has been found that although the molybdenum gross erosion peaks close to the separatrix, the net erosion peaks further away in the target plate. Three surfaces have been identified spectroscopically in C-Mod to be significant sources of molybdenum: the inner wall, the outer divertor and the antenna protection tiles. The inner wall is the only important source during limited plasma operation, while, in diverted discharges, the molybdenum generated there is very well screened by the plasma. In RF-heated diverted plasmas, it is believed that the antenna protection tiles are the source of most of the molybdenum that ends up in the core. The outer divertor can not be excluded as a contributor to the core molybdenum density but there are indications that it is often not the dominant source during RF heating. This result is significant since it is expected that the divertor target in the next generation fusion devices will be made primarily with a high-Z material. The study of boronization as a surface conditioning method which reduces the molybdenum source rates and core concentration has shown a varying effectiveness dependent on first wall surface location. The beneficial effects of boronization disappear rather fast for the outer divertor, last longer for the inner wall, with the antennas and plasma core benefiting the most. / by Dimitrios Pappas. / Ph.D.

Thesis (Sc. D.)--Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 1986. / MICROFICHE COPY AVAILABLE IN ARCHIVES AND SCIENCE / Bibliography: v.1, leaves 287-292. / by Victor Iannello. / Sc.D.

Thesis (S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 2004. / Includes bibliographical references (p. 59-60). / This electronic version was submitted by the student author. The certified thesis is available in the Institute Archives and Special Collections. / MIDAS is the minor incident decision analysis software that acts as an advisory tool for plant decision makers and operators to analyze the available decision alternatives for resolving minor incidents. The minor incidents dealt with in this thesis include non- safety related component failure, equipment maintenance, inspection or testing. MIDAS implements the risk-informed decision analysis methodology that uses multi- attribute utility theory (MAUT) and formal decision-making models that was developed for nuclear power plants. MIDAS integrates the theory, practical models and the graphical user interfaces for analysts to quickly obtain the insight regarding the performance of decision options and driving factors. To be able to deal with the inherent diversity of scenarios and decision options, a well-defined option models and modular calculation structure were constructed in MIDAS. In addition, MIDAS provides the functions of performing sensitivity and uncertainty analyses to take into account the inherent model and parameter uncertainties in decision option evaluation. Two case studies are performed to demonstrate the application of MIDAS in nuclear power plant risk-informed incident management. The insight obtained from the analysis results of case studies reveals that for nuclear power plant incident management, risk usually is not the most important concern. Cost and external attention are usually the dominant deciding factors in decision-making. However, in fact, the safety performance of each option is reflected in terms of the cost and external attention. / by Tze-Chieh Horng. / S.M.

Thesis (S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 2004. / Includes bibliographical references (p. 75-76). / One of the major challenges to the successful deployment of new nuclear plants in the United States is the regulatory process, which is largely based on water-reactor technology. While ongoing and expected efforts to license new LWR designs are based primarily on current regulations, guidance, and past licensing experience, the pre- application review of the gas-cooled Pebble Bed Modular Reactor (PBMR) has shown that efforts are being made to provide additional "risk-informed" improvements to the licensing process. The work presented herein was completed as part of the DOE Nuclear Energy Research Initiative project on a new "highly risk-informed" design and regulatory process. This study concentrates on the application of the risk-informed principles to a new plant design such as the PBMR. The purpose of this work is to provide selected examples of PRA applications, including development of Master Logic Diagram (MLD) for the selection of accident-initiating events and safety classification systems and components, as well as evaluating the risk significance of design features responsible for preventing and mitigating accidents. An implementation example chosen for a detailed analysis concentrates on the investigation of potential failure modes where performance of the novel design features such as water-based Reactor Cavity Cooling System (RCCS) is critical for the plant safety. The role and importance of the PBMR safety features is investigated by evaluating the risks from the most important external event: earthquake with a subsequent loss of offsite power. / (cont.) The scope includes specifying design configurations and using PRA techniques to evaluate the design, then iterating with subsequent design changes that improve the overall level of safety and system reliability. The viability of the new risk-informed process is demonstrated. Technical results, consistent with the known inherent safety features of such a reactor design, indicate that a pressure-tight containment similar to those for today's operating reactors may not be required for the PBMR. / by Marina D. Savkina. / S.M.

Thesis (Sc. D.)--Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 2002. / Includes bibliographical references (leaves 192-194). / The growing need for improved project management technique points to the usefulness of a knowledge-base advisory system to help project managers understand current and future project status and optimize decisions based upon the project performances. The work here demonstrates the framework of an advisory system with improved ability in project management. Based upon the literature survey and discussion with relevant experts, the Bayesian Belief Network (BBN) approach was selected to model the steam generator replacement proj ect management problem, where the situation holds inherently large uncertainty and complexities, since it has a superior ability to treat complexities, uncertainty management, systematic decision making, inference mechanism, knowledge representation and model modification for newly acquired knowledge. Two modes of advisory system have been constructed. As the first mode, the predictive mode has been developed, which can predict future project performance state probability distributions, assuming no intervening management action. The second mode is the advisory mode, which can identify the optimal action among alternatives based upon the expected net benefit values that are incorporating two important components: 1) expected immediate net benefits at post-action time, and 2) the expected long term benefit (or penalty) at scheduled project completion time. During the work, new indices for important variables have been newly developed for effective and efficient project status monitoring. With application of developed indices to the advisory system, the long term benefit (or penalty) found to be the most important factor in determining the optimal action by the project management during the decision / (cont.) making process and was confirmed by the domain experts. As a result, the effort has been focused on incorporating the long term benefit (or penalty) concept in order to provide more reliable and accurate advice to the project managers. In addition, in order to facilitate the communication between the BBN models and the users, an interface program has been developed using the Visual Basic language. / by Dohyoung Kim. / Sc.D.

Thesis (S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 2004. / Includes bibliographical references (leaves 132-133). / This thesis contributes to the hydride nuclear fuel project led by U. C. Berkeley for which MIT is to perform the thermal hydraulic and economic analyses. A parametric study has been performed to determine the optimum combination of lattice pitch, rod diameter, and channel shape-further referred to as geometry-for maximizing power given specific transient conditions for pressurized water reactors (PWR) loaded with either U02 or UZrH1.6 fuel. Several geometries have been examined with the VIPRE subchannel analysis tool along with MATLAB scripts previously developed to automate VIPRE execution. The transients investigated were a large break loss of coolant accident (LBLOCA), am overpower transient, and a complete loss of flow accident. The maximum achievable power for each geometry is defined as the highest power that can be sustained without exceeding any of the steady state or transient limits. The limits were chosen based on technical feasibility and safety of the reference core and compared with the final safely analysis report (FSAR) of the reference core, the South Texas Project Electric Generating Station (STPEGS), whenever possible. This analysis was performed for two separate pressure drop limits of 29 and 60 psia for both a square array with grid spacers and a hexagonal array with wire wraps. / (cont.) The square core geometry sustaining the highest power (4820.0 MW) for both the hydride and oxide fueled has a pitch of 9.0 mm and a rod diameter of 6.5 mm and was limited by the complete loss of flow accident. Both of these maximum power geometries occurred at the 60 psia pressure drop case. The maximum power of the 29 psia pressure drop case (4103.9 MW) for both fuel types occurred at a pitch of 9.7 mm and a rod diameter of 6.5 mm. The maximum power for the hexagonal arrayed cores occurred at the same hydrogen to heavy metal ratio as the square cores. The hydride fueled core power (5123.2 MW) was limited by the overpower transient while the oxide fueled core power (4996.1 MW) was limited by the overpower transient. The pressure drop constraint was not limiting for either fuel type for either pressure drop case for the wire wrapped cores. / by Jarrod Michael Trant. / S.M.

Thesis (Ph. D.)--Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 2005. / Includes bibliographical references (p. 211-226). / Edge transport barriers (ETBs) in tokamak plasmas accompany transitions from low confinement (L-mode) to high confinement (H-mode) and exhibit large density and temperature gradients in a narrow pedestal region near the last closed flux surface (LCFS). Because tokamak energy confinement depends strongly on the boundary condition imposed by the edge plasma pressure, one desires a predictive capability for the pedestal on a future tokamak. On Alcator C-Mod, significant contributions to ETB studies were made possible with edge Thomson scattering (ETS), which measures profiles of electron temperature (20 Te[eV] 800) and density (0.3 ne[1020m-3] ' 5) with 1.3-mm spatial resolution near the LCFS. Profiles of Te, ne, and Pe = neTe are fitted with a parameterized function, revealing typical pedestal widths A of 2-6mm, with ATe Ane, on average. Pedestals are examined to determine existence criteria for the enhanced D, (EDA) H-mode. A feature that distinguishes this regime is a quasi-coherent mode (QCM) near the LCFS. The presence or absence of the QCM is related to edge conditions, in particular density, temperature and safety factor q. Results are consistent with higher values of both q and collisionality v* giving the EDA regime. Further evidence suggests that increased Vpel may favor the QCM; thus EDA may have relevance to low-v* reactor regimes, should sufficient edge pressure gradient exist. / (cont.) Scaling studies of pedestal parameters and plasma confinement in EDA H-modes varied operational parameters such as current Ip and L-mode target density ne,L. At fixed plasma shape, widths show little systematic variation with plasma parameters. Scalings are however determined for pedestal heights and gradients. The Pe pedestal height and gradient both scale as I, similar to scalings found on other tokamaks, though with differing pedestal-limiting physics. It is seen that the density pedestal value ne,PED scales linearly with Ip, and more weakly with h,L, indicating that neutral fueling plays a relatively limited role in setting H-mode density. Plasma stored energy scales in a linear fashion with the Pe pedestal, such that empirical confinement scalings are affected by edge pedestal scalings. Empirical determination of neutral density and ionization source was made across the pedestal region, enabling inference of neutral gradient scale length Lo and effective diffusivity Def. The Def well is comparable in width to the pedestal, and Lo tends to be less than Ane. Computation of Lo in discharges with varying ne,L yields a similar result, suggesting that A,, is generally set by the ETB extent and not neutral penetration length. Puffing gas into an existing H-mode edge yields no significant change in the values of ne,PED, Vne, which is qualitatively consistent with simulations using a coupled fluid-kinetic neutral model. Experiment and modeling indicate the importance of thermal equilibration of neutrals with ions, particularly in high density collisional) plasmas. / by Jerry W. Hugues, Jr. / Ph.D.

Thesis (M.S.)--Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 1988. / Includes bibliographical references. / by Thomas W. Seubert. / M.S.

Thesis (M.S.)--Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 1984. / MICROFICHE COPY AVAILABLE IN ARCHIVES AND SCIENCE. / Includes bibliographical references. / by Frank Michael Gerard Wong. / M.S.

Thesis (M.S.)--Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 1990. / Includes bibliographical references (leaves 89-90). / by Everett Lee Redmond II. / M.S.