Thesis (S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 1997. / Includes bibliographical references (leaves 52-54). / A model has been developed to predict the pressure generation and concentrations of oxygen and hydrogen gas in a Multi-Canister Overpack (MCO) containing metallic uranium N Reactor fuel during staging, drying, conditioning, and storage at the Hanford Site. Ass0ciated uncertainties with each parameter were also calculated. The mechanisms that were included in the analysis were: -- Uranium Corrosion in which oxygen is consumed, or water is consumed and hydrogen is produced, -- Uranium Hydriding in which hydrogen is consumed, and -- Radiolysis in which water is consumed and hydrogen and oxygen are produced. The model characterizes the evolution of gases inside the MCO for the three ,egimes in which the fuel will be staged or stored. Prior to treatment the fuel will be immersed in water in the MCO. Proceeding the first treatment step, the MCO will contain no free water, but water vapor will contribute to a moist atmosphere. After the last treatment step, the inside of the MCO will be dry with only tightly-bound chemisorbed water remaining associated with sludge that will be present with the fuel. The model shows that for likely conditions inside the MCO, the container will not pressurize during its expected service life. The model also shows the effects that varying parameters have on the final pressure. / by Lisa Stiles Shell. / S.M.

Thesis (Ph. D.)--Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 1997. / Includes bibliographical references (p. 329-338). / Two complementary approaches were used to characterize transport on the Alcator C­Mod tokamak. The first was an empirical analysis of the scaling of transport with P*, the ion Larmor radius normalized to the plasma size. The second was a comparison of the transport predictions from the IFS-PPPL model of ion temperature gradient (ITG) driven turbulence to observations on C-Mod. The P* scaling experiments on C-Mod extend the range of plasma parameters over which the dimensionless scaling approach has been tested in both magnetic field ( to 8 T) and density (to (ne ) = 3.8 x 1020 /m3) ... / by Jeffrey Marc Schachter. / Ph.D.

Thesis (Ph.D.)--Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 1984. / Bibliography: leaves 284-290. / by Apostolos Efthimiadis. / Thesis (Ph.D.)--Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 1984.

Thesis (Ph. D.)--Massachusetts Institute of Technology, Dept. of Nuclear Engineering, February 2004. / Includes bibliographical references (p. 249-262). / This electronic version was submitted by the student author. The certified thesis is available in the Institute Archives and Special Collections. / The once through nuclear fuel cycle adopted by the majority of countries with operating commercial power reactors imposes a number of concerns. The radioactive waste created in the once through nuclear fuel cycle has to be isolated from the environment for thousands of years. In addition, plutonium and other actinides, after the decay of fission products, could become targets for weapon proliferators. Furthermore, only a small fraction of the energy potential in the fuel is being used. All these concerns can be addressed if a closed fuel cycle strategy is considered offering the possibility for partitioning and transmutation of long lived radioactive waste, enhanced proliferation resistance, and improved utilization of natural resources. It is generally believed that dedicated advanced reactor systems have to be designed in order to perform the task of nuclear waste transmutation effectively. The development and deployment of such innovative systems is technically and economically challenging. In this thesis, a possibility of constraining the generation of long lived radioactive waste through multi-recycling of Trans-uranic actinides (TRU) in existing Light Water Reactors (LWR has been studied. Thorium based and fertile free fuels (FFF) were analyzed as the most attractive candidates for TRU burning in LWRs. Although both fuel types can destroy TRU at comparable rates (about 1150 kg/GWe-Year in FFF and up to 900 kg/GWe-Year in Th) and achieve comparable fractional TRU burnup (close to 50a/o), the Th fuel requires significantly higher neutron moderation than practically feasible in a typical LWR lattice to achieve such performance. / (cont.) On the other hand, the FFF exhibits nearly optimal TRU destruction performance in a typical LWR fuel lattice geometry. Increased TRU presence in LWR core leads to neutron spectrum hardening, which results in reduced control materials reactivity worth. The magnitude of this reduction is directly related to the amount of TRU in the core. A potential for positive void reactivity feedback limits the maximum TRU loading. Th and conventional mixed oxide (MOX) fuels require higher than FFF TRU loading to sustain a standard 18 fuel cycle length due to neutron captures in Th232 and U238 respectively. Therefore, TRU containing Th and U cores have lower control materials worth and greater potential for a positive void coefficient than FFF core. However, the significantly reduced fuel Doppler coefficient of the fully FFF loaded core and the lower delayed neutron fraction lead to questions about the FFF performance in reactivity initiated accidents. The Combined Non-Fertile and UO2 (CONFU) assembly concept is proposed for multi- recycling of TRU in existing PWRs. The assembly assumes a heterogeneous structure where about 20% of the UO2 fuel pins on the assembly periphery are replaced with FFF pins hosting TRU generated in the previous cycle. The possibility of achieving zero TRU net is demonstrated. The concept takes advantage of superior TRU destruction performance in FFF allowing minimization of TRU inventory. At the same time, the core physics is still dominated by UO2 fuel allowing maintenance of core safety and control characteristics comparable to all-UO2. / by Evgeni Shwageraus. / Ph.D.

Thesis (Ph. D.)--Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 2001. / Includes bibliographical references (p. 153-156). / This is a study of fault detection for nuclear reactor systems. Basic concepts are derived from fundamental theories on system observers. Different types of fault- actuator fault, sensor fault, and system dynamics fault can be detected and localized by studying the asymptotic response of an error signal constructed from the system inputs, system outputs, and observer outputs. False alarm and failure to detect a fault are two decision errors when noise is considered. The goal here is to achieve a reasonable compromise. The two types of decision errors can be characterized by their respective first hitting time of a decision threshold. This in turn is dependent on the design of the observer and the decision rule. Costs corresponding to these two types of decision error are defined by cost functions that are in turn constructed based on experience and knowledge of the system operation. A method has been developed in this research to find an optimal design of the observer, the design of a frequency-dependent output filter, and a decision rule that could achieve the desired economic goals. This technique is applied to nuclear reactor systems and simulations are carried out. The one-group linear nuclear reactor model is used in the observer. The system is modeled by a one-group linear model and by a six-group non-linear model. Results show that this fault detection method can not only detect a fault but also localize it at the same time by constructing specially targeted fault detection filters. These fault detection filters are robust against measurement noise and modeling errors. / by Qing Li. / Ph.D.

Thesis (Ph. D.)--Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 1996. / Includes bibliographical references (leaves 91-92). / by Yang Zhang. / Ph.D.

Thesis (M.S.)--Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 1963. / Includes bibliographical references (leaves [120]-[122]). / by Joseph Harrington, III. / M.S.

Thesis (M.S.)--Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 1986. / MICROFICHE COPY AVAILABLE IN ARCHIVES AND SCIENCE. / Bibliography: leaves 75-76. / by Gordon M. Shedd. / M.S.

Thesis (S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 2003. / Includes bibliographical references (p. 126-128). / This electronic version was submitted by the student author. The certified thesis is available in the Institute Archives and Special Collections. / The feasibility of using various uranium-free fuels for plutonium incineration in present light water reactors is investigated. Two major categories of inert matrix fuels are studied: composite ceramic fuel particles dispersed in another ceramic matrix (CERCER) and ceramic fuel particles dispersed into a metallic matrix (CERMET). In the category of CERCER, the current world wide research effort has been focused on three matrix candidates: (1) Spinel (MgAl2O4); (2) CeO2, and (3) MgO. In contrast, there are still no emerging commonly accepted matrix candidates for a CERMET. The fuel may consist of plutonium, minor actinides (MA), or both which are termed trans-uranium (TRU) fuel. The transmutation rate and the transmuted fraction of initial loadings are calculated using CASMO-4. Different inert matrix fuels have similar burning abilities in terms of how much and how fast the Pu, MA or TRU can be burned, and they are all superior to the mixed UO2-PuO2 (MOX) fuel. From this point of view, there is no good reason to favor one inert matrix over another. The burning rates in terms of kg/(GWe-Year) of different inert matrix fuels are quite stable with regard to changing the moderation level (or H/HM ratio) in the core. Changing initial loadings and changing power densities can not result in large change in the burned percentage of initial loadings and burning rate. Lack of U-238 and the neutronic characteristics of plutonium lead to degradation of safety related kinetic parameters. It is found that various inert matrix fuels have similar values for the Doppler coefficient, moderator temperature coefficient, void coefficient, boron worth and effective delayed neutron fraction [beta]eff. But their Doppler coefficients are much smaller than those of MOX and UO2 fuels. Both inert matrix fuels and MOX fuel have a much smaller effective delayed neutron fraction than UO2 fuel. IMF fuel's value is smaller than that of UO2 fuel, but close to that of MOX. The void coefficient is also a potential problem. The coolant void reactivity worth becomes positive if the void fraction reaches 80 percent at beginning of life for boron concentration of 1500 ppm. This is confirmed by MCNP- 4C and modified CASMO-4 calculations. The situation is generally much worse at BOL than at EOL. This is of concern during loss of coolant accidents. Two options are explored in order to improve safety coefficients: 1. Adding fertile materials into IMF fuel pins; 2. Adding fertile fuel pins into IMF fuel assemblies. UO2 and ThO2 are used as fertile additives. In option 1 it is found that adding UO2 will result in a worse degradation of burning ability. However, adding UO2 provides a better Doppler coefficient than adding ThO2. Adding about 20 w/o of UO2 achieves a BOL Doppler coefficient and other safety coefficients comparable to the traditional UO2 fuel. Yet, the fuel still has a much better burning percentage than MOX fuel for Pu (47.5% versus 13.2%) and for MA (36.1% versus 19.0%), and a much better burning rate for Pu (834 versus 285 kg/GWe-Year) and for MA (98 versus 63 kg/GWe-Year). In option 2 it is found that in order to reach the same level of safety coefficients, the burning rate of the minor actinides becomes comparable to that of MOX: 68.7 versus 63.3 kg/GWe-Year. Thus the option of adding fertile material into the fuel pins is preferable over a heterogeneous assembly option if fast burning of minor actinides is favored. / by Yi Zhang. / S.M.

Thesis (Ph. D.)--Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 2004. / Page 242 blank. / Includes bibliographical references (p. 238-241). / The objective of this study is to improve quality and reliability of safety-critical software in the nuclear industry. It is accomplished by focusing on the following two areas: Formulation of a standard extensive integrated software testing strategy for safety-critical software, and Development of systematic test-based statistical software reliability quantification methodologies. The first step to improving the overall performance of software is to develop a comprehensive testing strategy, the gray box testing method. It has incorporated favorable aspects of white box and black box testing techniques. The safety-critical features of the software and feasibility of the methodology are the key drivers in determining the architecture for the testing strategy. Monte Carlo technique is applied to randomly sample inputs based on the probability density function derived from the specification of the given software. Software flowpaths accessed during testing are identified and recorded. Complete nodal coverage testing is achieved by automatic coverage checking. It is guaranteed that the most popular flowpaths of the software are tested. / The second part of the methodology is the quantification of software performance. Two Bayesian based white box reliability estimation methodologies, nodal coverage- based and flowpath coverage-based, are developed. The number of detected errors and the failure-free operations, the objective and subjective knowledge of the given software, and the testing and software structure information are systematically incorporated into both reliability estimation approaches. The concept of two error groups in terms of testability is initiated to better capture reliability features of the given software. The reliability of the tested flowpaths of the software and that of the untested flowpaths can be updated at any point during testing. Overall software reliability is calculated as a weighted average of the tested and untested parts of the software, with the probability of being visited upon next execution as the weight of each part. All of the designed testing and reliability estimation strategies are successfully implemented and automated via various software tools and demonstrated on a typical safety-critical software application. / by Yi Zhang. / Ph.D.